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Attachment 6, LaSalle, Unit 2, Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex
	ML092810281
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Site:	LaSalle
Issue date:	08/31/2009
From:	; AREVA NP
To:	; Office of Nuclear Reactor Regulation
References
	RS-09-133 ANP-2843(NP), Rev. 1
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{{#Wiki_filter:ATTACHMENT 6 AREVA NP Inc. Affidavit and Non-Proprietary Version of Attachment 3 AFFIDAVIT COMMONWEALTH OF VIRGINIA )) ss.CITY OF LYNCHBURG )1. My name is Gayle F. Elliott. I am Manager, Product Licensing, for AREVA NP Inc. and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by AREVA NP to determine whether certain AREVA NP information is proprietary.

I am familiar with the policies established by AREVA NP to ensure the proper application of these criteria.3. I am familiar with the AREVA NP information contained in the report ANP-2843(P), Revision 1, entitled "LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex," dated August 2009 and referred to herein as "Document." Information contained in this Document has been classified by AREVA NP as proprietary in accordance with the policies established by AREVA NP for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA NP and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure.

The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information." 6. The following criteria are customarily applied by AREVA NP to determine whether information should be classified as proprietary: (a) The information reveals details of AREVA NP's research and development plans and programs or their results.(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA NP.(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA NP in product optimization or marketability.(e) The information is vital to a competitive advantage held by AREVA NP, would be helpful to competitors to AREVA NP, and would likely cause substantial harm to the competitive position of AREVA NP.The information in the Document is considered proprietary for the reasons set forth in paragraphs 6(b) and 6(c) above.7. In accordance with AREVA NP's policies governing the protection and control of information, proprietary information contained in this Document have been made available, on a limited basis, to others outside AREVA NP only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. AREVA NP policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.SUBSCRIBED before me this____day of August 2009.Sherry L. McFaden NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 10/31/10 Reg. # 7079129 Notafy Publo Commonwealth of V9rglnlo 7079129 My Commlieln Expires Oct 31. 201t A ANP-2843(NP Revision'k: LaSalle Unit 2 Nuclear Power Station Spent Fue Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Borafle>August 2009)I AREVA NP Inc.ANP-2843(NP)

Revision 1 LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex sip AREVA NP Inc.ANP-2843(NP) Revision 1 Copyright © 2009 AREVA NP Inc.All Rights Reserved AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page i Nature of Changes Item Page Description and Justification 1 2 3 4 2-4 2-5 4-4,6-13 5-1 Punctuation corrected 4 0C added for clarification Proprietary markings removed from insert parameters Space added AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page ii Contents 1.0 Introduction .................................................................................................................... 1-1 2.0 Sum mary ........................................................................................................................ 2-1 3.0 Criticality Safety Design Criteria ..................................................................................... 3-1 4.0 Fuel and Storage Array Description .......................................................................... 4-1 4.1 Fuel Assem bly Design .................................................................................... 4-1 4.2 Fuel Storage Racks ........................................................................................ 4-1 5.0 Calculation Methodology ........................................................................................... 5-1 5.1 Area of Applicability .................................................................................................. 5-2 6.0 Criticality Safety Analysis ............................................................................................... 6-1 6.1 Geometry Model ................................................................................................. 6-1 6.2 Definition of REBOL Lattices .............................................................................. 6-1 6.3 Storage Array Reactivity ..................................................................................... 6-3 6.4 Uncertainties ...................................................................................................... 6-4 6.5 Abnorm al and Accident Conditions .................................................................... 6-4 6.6 Determ ination of Maxim um Rack Assem bly k-eff ............................................... 6-6 6.7 Uniform vs. Distributed Enrichment Distributions ............................................... 6-7 6.8 Arrays of M ixed BW R Fuel Types ...................................................................... 6-7 6.9 Inaccessible Storage Locations .......................................................................... 6-8 6.10 Interfaces between Areas with Different Storage Conditions ............................. 6-8 7.0 Conclusions .................................................................................................................... 7-1 8.0 References ..................................................................................................................... 8-1 Appendix A Appendix B Appendix C Appendix D Sample CASMO-4 Input ............................................................................. A-1 Reactivity Comparison for Assemblies Used in the LaSalle R e acto rs .................................................................................................... ..B -1 KENO V.a Bias and Bias Uncertainty Evaluation ....................................... C-1 CASMO-4 Benchmarking for In-Rack Modeling .......................................... D-1 AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page iii Tables 2.1 Criticality Safety Limitations for ATRIUM-10 Fuel Assemblies Stored in the LaSalle Unit 2 Nuclear Power Station Spent Fuel Pool ............................................ 2-4 4.1 ATRIUM-10 Fuel Assembly Parameters ........................................................................ 4-3 4.2 Fuel Storage Rack Param eters ...................................................................................... 4-4 6.1 Summary of CASMO-4 Maximum Reactivity Results for the ATRIUM-10 F ue l A sse m b ly .............................................................................................................. 6 -10 6.2 Summary of KENO V.a Maximum In-Rack Reactivity for ATRIUM-10 Fuel ................. 6-12 6.3 Manufacturing Reactivity Uncertainties ........................................................................ 6-13 6.4 Evaluation for Inaccessible Storage Locations ........................................................... 6-14 Figures 2.1 ATRIUM-10 Reference Bounding Assembly .................................................................. 2-6 4.1 4.2 4.3 Representative ATRIUM -10 Fuel Assem bly .................................................................. 4-5 Calculational Model of Storage Cell ............................................................................... 4-6 Storage Rack with Inserts .............................................................................................. 4-7 AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel ANP-2843(NP) Storage Pool Criticality Safety Analysis with Revision 1 Neutron Absorbing Inserts and Without Boraflex Page iv Nomenclature BAF bottom of active fuel BOL beginning of life BWR boiling-water reactor CPR critical power ratio CW clock-wise EALF the energy of the average lethargy causing fission GWd energy unit, giga-watt-day k-eff effective neutron multiplication factor k. infinite lattice neutron multiplication factor LHGR linear heat generation rate PLR part-length fuel rod NRC Nuclear Regulatory Commission, U. S.REBOL reactivity-equivalent at beginning of life (fresh fuel, no Gd 2 0 3 , no fission products)TD theoretical density H/X atomic ratio of hydrogen (H) to fissile isotopes (X)AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page 1-1 1.0 Introduction This report presents the results of a criticality safety evaluation performed for the LaSalle Unit 2 Nuclear Power Station spent fuel storage pool assuming complete Boraflex degradation and the use of neutron absorbing inserts in each accessible storage cell. Reference 1 is the last criticality safety evaluation that was submitted for NRC review for the LaSalle Unit 2 spent fuel pool.In this report, a reference bounding assembly has been defined to bound the reactivity of all past and current fuel assembly types delivered to the LaSalle station (both Units 1 and 2). This reference bounding assembly is based on an AREVA NP Inc.* ATRIUMt-10 fuel assembly. This analysis demonstrates that with the reference bounding assembly, complete Boraflex degradation, and a neutron absorbing NETCO-SNAP-IN insert in each storage cell, the pool k-eff remains below the 0.95 k-eff acceptance criterion established by the NRC.* AREVA NP Inc. is an AREVA and Siemens company.t ATRIUM is a trademark of AREVA NP.AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel ANP-2843(NP) Storage Pool Criticality Safety Analysis with Revision 1 Neutron Absorbing Inserts and Without Boraflex Page 2-1 2.0 Summary Criticality analyses have been performed and are documented herein for the LaSalle Unit 2 spent fuel pool assuming no Boraflex and the presence of a NETCO-SNAP-IN insert in each accessible storage cell of the rack. The criticality analyses are based on the use of a reference fuel assembly design that is bounding of (i.e., more reactive than) all fuel designs used in Units 1 and 2 at the LaSalle station. The KENO V.a code was used for all calculations that do not require fuel depletion. The CASMO-4 code is used to compare lattice k. values at peak reactivity conditions and in defining the gadolinia manufacturing uncertainty. Benchmarking is included for both the KENO V.a and CASMO-4 codes.The calculations documented herein demonstrate that the ATRIUM-10 reference bounding assembly design has been selected to be more reactive, in an in-rack configuration without Boraflex and with the NETCO-SNAP-IN inserts, than any of the current or past fuel assembly designs used in the LaSalle reactors. These comparisons are based upon actual GE 8x8, ATRIUM-9, GEI4, ATRIUM 1OXM and ATRIUM-10 lattice geometries and enrichment distributions and the results are shown in Appendix B. This evaluation establishes that the fuel assemblies previously manufactured for use in the LaSalle reactors can be safely stored in the LaSalle Unit 2 spent fuel storage pool with NETCO-SNAP-IN inserts.The reference bounding assembly is defined with two U235 enrichment / gadolinia concentration zones. The bottom enrichment / gadolinia zone is divided into two separate axial zones by the ATRIUM-10 geometry transition at 96". This creates the 3 zones shown in Figure 2.1. Three REBOL lattices have been defined to represent the lattices of the reference bounding assembly in KENO calculations. The reactivity of the REBOL lattices have been increased to compensate for the uncertainties associated with defining these maximum reactivity lattices.This evaluation includes manufacturing uncertainties for the ATRIUM-1 0 fuel design and the fuel pool storage racks, code modeling uncertainties, reactivity increases due to accident or abnormal conditions, and one-sided tolerance multipliers to determine the 95/95 upper limit k-eff. The conditions and uncertainties assumed in this analysis are described in Section 6.This evaluation demonstrates that the reference ATRIUM-10 fuel assembly does not exceed an array k-eff of 0.95 in the LaSalle Unit 2 spent fuel storage pool without Boraflex, provided the AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel ANP-2843(NP) Storage Pool Criticality Safety Analysis with Revision 1 Neutron Absorbing Inserts and Without Boraflex Page 2-2 neutron absorbing insert depicted in Figure 4.2 has been installed in each accessible storage cell.As defined in Table 2.1, ATRIUM-10 fuel that contains equivalent or less enrichment and equivalent or higher Gd 2 0 3 concentrations in the fuel zones depicted in Figure 2.1 can be safely stored in the LaSalle Unit 2 spent fuel storage pool. In addition, ATRIUM-10 fuel that contains more enrichment and/or lower Gd 2 0 3 concentrations than the reference assembly design can be safely stored provided each zone of the assembly is less reactive than the corresponding zone of the reference assembly design. This can be established using the storage rack model in the CASMO-4 lattice physics code as described in Appendix A.This analysis considers unchanneled fuel assemblies as well as assemblies with the AREVA 100 mil fuel channel.* Additionally, there is no limitation for bundle orientation or position in the storage cell since these are accounted for in the analysis.To assure that the actual reactivity will always be less than the calculated reactivity, the following conservative assumptions have been made:* The results are based on a moderator temperature of 40C (39.2°F), which gives the highest reactivity for the fuel storage pool for a configuration assuming no Boraflex with NETCO-SNAP-IN inserts.* Fuel assemblies are assumed to contain the high reactivity reference bounding lattices for the entire length of the assembly, (natural uranium blankets are not modeled).* Each lattice in each fuel assembly in the array is assumed to be at its lifetime maximum reactivity level, (no credit is taken for assembly burnup).* The most limiting orientation or position of each assembly in its rack cell is accounted for in the analysis.* The analysis takes into account storage with or without fuel channels. (The array k-eff is higher with a fuel channel present).* Neutron absorption in fuel assembly structural components (spacerst, tie plates, etc) is neglectedY.

The maximum reactivity value includes all significant manufacturing and calculational uncertainties.
The AREVA advanced fuel channel and the AREVA 80 mil fuel channel are also acceptable.

t It is conservative to neglect the spacers because this spent fuel pool contains no soluble boron and the region around the fuel rods is under-moderated and neglecting the spacer places more water within the calculational model. In addition, the inconel springs are a stronger neutron absorber than water.The active fuel region repeats periodically in the vertical direction. Therefore, neutron absorption in upper and lower tie plates, fuel plenums, etc. is neglected. AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page 2-3 The reactivity of the REBOL lattices used in the KENO analysis have been designed to be at least 0.010 Ak more reactive than the reference bounding lattices they represent. This is more than the uncertainty associated with defining these maximum reactivity lattices.AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbina Inserts and Without Boraflex ANP-2843(NP) Revision 1 Paae 2-4 Table 2.1 Criticality Safety Limitations for ATRIUM-10 Fuel Assemblies Stored in the LaSalle Unit 2 Nuclear Power Station Spent Fuel Pool 1. ATRIUM-10 Fuel Configuration Parameter Nominal ATRIUM-10 Value Clad OD, in. 0.3957 Clad ID, in. 0.3480 Pellet Diameter, in. 0.3413 Rod Pitch, in. 0.510 Fuel Density % Theoretical 95.85 to 96.26 Water Rods Internal Channel 2. Fuel may be stored with or without fuel channels.3. Fuel Design Limitations for Enriched Lattices*The U235 enrichment and gadolinia concentration levels must meet the requirements specified below and shown graphically in Figure 2.1 (dimensions represent fuel column height above BAF).Above 126" Maximum Lattice Average Enrichment, wt% U-235 4.47 Minimum Number of Rods containing Gd 2 0 3 10 Minimum wt% Gd 2 0 3 in each Gd Rod 3.5 Below 126" t Maximum Lattice Average Enrichment, wt% U-235 4.57 Minimum Number of Rods containing Gd 2 0 3 10 Minimum wt% Gd 2 0 3 in each Gd Rod 6.0 Eight gadolinia rods must be loaded one row in from the edge of the lattice such that rows 2 and 9 and columns 2 and 9 each contain 2 gadolinia rods.4. ATRIUM-10 fuel assemblies which do not meet the limitations above may be stored in the LaSalle Unit 2 spent fuel pool provided the reactivity of any enriched lattice does not exceed the following in-rack k. values at any point during their lifetime. (The CASMO-4 storage rack model that must be used for this calculation is defined in Appendix A and the* These are the reference bounding lattices described on Page 6-2.This is actually two axial zones divided by the geometry of the ATRIUM-10 part-length rod transition at 96" above BAF.AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page 2-5 transition between top and bottom lattice geometries occurs at 96 inches from the bottom of the fueled length.)Zone Lattice Geometry Distance from BAF Max. in-rack k. (4 0 C)3 AlOT (83 rods) 126" to 149" 0.9185 2 AlOT (83 rods) 96" to 126" 0.8869 1 A10B (91 rods) 0" to 96" 0.8843 5. The spent fuel storage rack design parameters and dimensions are as defined in Reference 4, and a general description of the NETCO-SNAP-IN inserts is provided in Reference 5.AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page 2-6 149.0" AlOT-4.47L1 0G3.5 126.0" Al0T-4.57L1 0G6.0 96.0" AlOB-4.57L1 0G6.0 0.0"1 Figure 2.1 ATRIUM-10 Reference Bounding Assembly AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page 3-1 3.0 Criticality Safety Design Criteria The criticality safety design criteria defined in the following documents are applicable for this LaSalle Unit 2 Nuclear Power Station spent fuel storage facility evaluation: A. Subsection B.4 of 1 OCFR 50.68, (Criticality Accident Requirements), (Reference 6).B. Section 9.1.1 (Fresh and Spent Fuel Storage and Handling) of the Standard Review Plan (Reference 7).C. ANSI/ANS American National Standard 57.2-1983 (Design Requirements for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants) issued by the American Nuclear Society (Reference 8).*D. ANSI/ANS American National Standard 8.17-1984 (Criticality Safety Criteria for the Handling, Storage and Transportation of LWR Fuel Outside Reactors) issued by the American Nuclear Society, January 1984 (Reference 9).E. "OT Position for the Review and Acceptance of Spent Fuel Storage and Handling Applications," issued by the NRC in 1978 and amended in 1979 (Reference 10).F. "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," issued by the NRC in 1998 (Reference 11).These documents define the assumptions and acceptance criteria used in this evaluation. In descending order (from A to F), these documents go from "least" to "most" detail relative to explicitly defining what needs to be addressed in the criticality safety evaluation. In general, the criticality safety acceptance criterion applicable to this evaluation is as defined by Section 9.1.1 of the Standard Review Plan (Reference 7):...the k-eff will not exceed 0.95 for all normal and credible abnormal conditions. This is consistent with requirements in the LaSalle FSAR and Technical Specifications.

ANSI/ANS 57.1 and 57.3 are endorsed in combination with ANSI/ANS 57.2 in item B. ANSI/ANS 57.1 and 57.3 are not cited here because they do not apply to spent fuel pool criticality.

AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel ANP-2843(NP) Storage Pool Criticality Safety Analysis with Revision 1 Neutron Absorbing Inserts and Without Boraflex Page 4-1 4.0 Fuel and Storage Array Description LaSalle Units 1 and 2 have loaded four different product lines-GE 8x8 fuel, ATRIUM-9 fuel, GE14 fuel, and ATRIUM-10 fuel. The ATRIUM-10 fuel product line is the fuel currently being loaded in reload quantities in both LaSalle reactors. All four of these designs are stored in the LaSalle Unit 2 spent fuel storage pool. In an in-rack configuration assuming no Boraflex and NETCO-SNAP-IN inserts, the reference ATRIUM-10 design has a higher reactivity than all previously loaded fuel assembly designs. Appendix B provides information from which this conclusion can be made. As such, the ATRIUM-10 reference bounding assembly design forms the basis for demonstrating that the maximum k-eff of the spent fuel pool storage array without Boraflex with NETCO-SNAP-IN inserts remains less than 0.95.4.1 Fuel Assembly Design The ATRIUM-10 fuel assembly is a 10x10 fuel rod array with an internal square water channel offset in the center of the assembly (taking the place of nine fuel rod locations). The assembly contains part-length fuel rods (PLR); therefore, a "top" lattice geometry will apply above the PLR boundary and a "bottom" lattice geometry will apply below the PLR boundary. The ATRIUM-10 mechanical design parameters are summarized in Table 4.1. A representation of the ATRIUM-10 assembly design is depicted in Figure 4.1. The ATRIUM-10 fuel in the LaSalle Nuclear Power Station has used and will use the standard 100 mil fuel channel design.4.2 Fuel Storage Racks The spent fuel storage rack dimensions and details are shown in Reference

4. The key rack assembly dimensions and tolerances are listed in Table 4.2. The fuel pool storage cell with ATRIUM-10 fuel has been modeled in CASMO-4 as shown in Figure 4.2 with small variations in KENO V.a. Each rack consists of an array of stainless steel boxes with a separation of 0.075" between each box wall. Originally this separation was filled with a layer of Boraflex material;however, for this analysis it is assumed that the Boraflex has been removed and is now replaced with water.For this evaluation, a chevron shaped neutron absorbing insert (NETCO-SNAP-IN) is modeled in each of the storage cells (see the general description in Reference 5). These inserts will extend over the full length of the fueled zone and will maintain the same orientation in each AREVA NP Inc.

AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel ANP-2843(NP) Storage Pool Criticality Safety Analysis with Revision 1 Neutron Absorbing Inserts and Without Boraflex Page 4-2 storage cell. Based on the insert configuration of Figure 4.3, peripheral storage cells on the north and east sides of the storage pool will not be completely surrounded by four wings of the absorbing insert. In the actual Unit 2 pool configuration, there will also be a minimal number of peripheral cells on all sides of the storage pool that will not be completely surrounded by four wings of the absorbing insert due to geometric layout and inaccessible storage locations. AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page 4-3 Table 4.1 ATRIUM-10 Fuel Assembly Parameters Fuel Assembly Fuel Rod Array 10x10 Fuel Rod Pitch, in. 0.510 Number of Fuel Rods Per Assembly 91 Water Channel 1 Fuel Rods Fuel Material U0 2 Pellet Density, % of Theoretical 96.26*Pellet Diameter, in. 0.3413 Pellet Void Volume, %Enriched U02 1.21 Cladding Material Zircaloy-2 Cladding OD, in. 0.3957 Cladding ID, in. 0.3480 Internal Water Channel Outside Dimension, in.Inside Dimension, in.Channel Material Fuel Channel (standard 100 mil)t Outside Dimension, in.Inside Dimension, in.Channel Material Fuel Column Lengths Distance from the bottom of the fuel to the top of the fuel in the part length fuel rods, in.Total Fueled Length, in.1.378 1.321 Zircaloy 2 or Zircaloy-4 5.478 5.278 Zircaloy-2, Zircaloy-4, or Zirc-BWR 96.0 149.0* Criticality safety analysis is valid for nominal pellet densities between 95.85% and 96.26% TD.t Depending on pellet L/D, the pellet void volume can vary. A nominal value of 1.2% was assumed for the criticality safety analysis. Variations of the void volume are not significant relative to impact on storage array criticality safety. (Use of chamfered pellets with higher void volumes are also acceptable) The conclusions in this report are equally valid for fuel channels that may differ. Hence, conclusions remain valid for other fuel channel types, e.g., advanced channels etc. (See discussion about fuel channels in Section 6.2).AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel ANP-2843(NP) Storage Pool Criticality Safety Analysis with Revision 1 Neutron Absorbing Inserts and Without Boraflex Page 4-4 Table 4.2 Fuel Storage Rack Parameters Parameter Insert, B-10 areal density, g/cm 2 Insert wing thickness, in.Material Insert modeled wing length, in.Storage cell Inside Dimension, in.Inner rack box wall thickness, in.Box material Original Boraflex thickness, in.Material Nominal rack cell pitch, in.Value 0.0086 minimum 0.065 +/- 0.005 Aluminum and B-10 5.98*6.00 +/- 0.02 0.090 +/- 0.009 Stainless steel 0.075 +/- 0.007 Originally Boraflex, now modeled as water 6.255 [ I* Value used in the KENO model. 6.00" was used in the CASMO-4 model which requires the insert wing to extend to the inside wall of the fuel storage cell.AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page 4-5 Tie Plate tod bly Channel ULTRAFLOW Spacer J-Partial Length Fuel Rod Assembly ower Tie Plate Assembly Figure 4.1 Representative ATRIUM-10 Fuel Assembly (Assembly length and number of spacers has been reduced for pictorial clarity.)AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page 4-6 NEUTRON ABSORBING INSERT NOT TO SCALE 6.255" LATTICE SPACING PERIODIC BOUNDARY CONDITIONS AT CENTERLINE OF WATER (4 sides)Figure 4.2 Calculational Model of Storage Cell AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page 4-7 Neutron Absorbing Insert Stainless Steel (North)Not to Scale Figure 4.3 Storage Rack with Inserts AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel ANP-2843(NP) Storage Pool Criticality Safety Analysis with Revision 1 Neutron Absorbing Inserts and Without Boraflex Page 5-1 5.0 Calculation Methodology The spent fuel storage criticality safety evaluation is performed with the KENO V.a Monte Carlo code, which is part of the SCALE 4.4a Modular Code System (Reference 2). The ENDF/B-V, 44 energy group data library is used by the SCALE driver module CSAS25, which uses modules BONAMI-2 and NITAWL to perform spatial and energy self-shielding adjustments of the cross sections for use in KENO V.a. AREVA has benchmarked KENO V.a in accordance with NUREG/CR-6698 (Reference

3) using critical experiments related to the storage of fuel assemblies in water -including neutron absorbing materials such as stainless steel and BORAL.For applications using the 44 energy group data libraries, the KENO V.a bias and standard deviation are 0.00542 and 0.00511, respectively (see Appendix C).KENO V.a is run on the AREVA NP scientific computer cluster using the Linux operating system. The hardware and software configurations are governed by AREVA NP procedures to ensure calculational consistency in licensing applications.

The code modules are installed on the system and the installation check cases are run to ensure the results are consistent with the installation check cases that are provided with the code. The binary executables are put under configuration control so that any changes in the software will require re-certification. The hardware configuration of each machine in the cluster is documented so that any significant change in hardware or operating system that could result in a change in results is controlled. In the event of such a change in hardware or operating system, the hardware validation suite is rerun to confirm that the system still performs as it did when the code certification was performed. In this analysis the SCALE 4.4a code system is employed to:* Calculate Dancoff coefficients

Calculate absolute k-effective results for the LaSalle Unit 2 spent fuel pool* Evaluate accident conditions, alternate loading conditions, and manufacturing tolerance conditions The CASMO-4 code is used when conditions require fuel and gadolinia depletion.

CASMO-4 is a multigroup, two-dimensional transport theory code with an in-rack geometry option where typical storage rack geometries can be defined on an infinite lattice basis. This code is used for fuel depletion and relative reactivity comparisons in a manner that is consistent with AREVA's AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel ANP-2843(NP) Storage Pool Criticality Safety Analysis with Revision 1 Neutron Absorbing Inserts and Without Boraflex Page 5-2 NRC approved CASMO-4 / MICROBURN-B2 methodology (Reference 12). CASMO-4 has been approved at LaSalle for BWR calculations and is included as a methodology reference (via Reference 1.2) in Section 5.6.5.B of the LaSalle Technical Specifications. The CASMO-4 computer code is controlled by AREVA procedures and the version used in this analysis meets the requirements of Reference 12.In this analysis CASMO-4 is employed to:* Perform in-core isotopic depletion at [ ] void history levels for fuel lattices.* Perform in-rack k. assessments to identify the lattices with maximum reactivity. a Define lattices for a reference bounding assembly that represent the maximum reactivity condition supported by the analysis.* Define the reactivity equivalent, beginning-of-life (REBOL) lattices with fresh fuel and no gadolinia, for the subsequent KENO V.a base case criticality calculations. Note that for the REBOL lattices, the U-235 content is manually adjusted upward until the REBOL k. is at least 0.01 Ak greater than the lattices of the reference bounding assembly. This 0.01 Ak is used to account for calculational and depletion uncertainties of the CASMO-4 code as discussed in Appendix D.Evaluation of the manufacturing uncertainty for gadolinia content. This is needed since a lower gadolinia concentration will deviate from the nominal case more near peak reactivity than it will at beginning of life (i.e., in a REBOL assembly). 5.1 Area of Applicability Table C.6 in Appendix C shows the ranges of key parameters represented in the KENO V.a benchmark analysis. Parameters such as rectangular lattices of zircaloy clad U02 fuel rods in a pool of water with stainless steel and boron are sufficiently general to not require comparison. The remaining parameters are compared in the following table and show that the KENO V.a portion of this analysis has been performed within the range of experimental conditions used in the KENO V.a benchmark. AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page 5-3 Parameter Enrichment (wt% U-235)Pitch (cm)H/X ratio Energy of the Average Lethargy Causing Fission (eV)Benchmark Values Values in this Analysis 2.46 to 9.83 2.66 to 4.57 1.04 to 2.64 1.27 to 1.31 17.4 to 473 250 to 350 0.11 to 2.51 0.19 to 0.26 For the CASMO-4 qualification, ATRIUM-10 fuel lattices were modeled using the LaSalle fuel storage rack geometry. Therefore, the CASMO-4 calculations performed for this evaluation are within the area of applicability of the comparisons shown in Appendix D.AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel ANP-2843(NP) Storage Pool Criticality Safety Analysis with Revision 1 Neutron Absorbing Inserts and Without Boraflex Page 6-1 6.0 Criticality Safety Analysis The criticality safety evaluation uses a reference bounding assembly comprised of two top and one bottom geometry reference bounding lattices* to demonstrate that the upper limit k 9 5/9 5 k-eff for the LaSalle Unit 2 Nuclear Power Station spent fuel pool can be met. These evaluations include the worst credible conditions and uncertainties as defined in the references documented in Section 3.0. The reference bounding ATRIUM-10 bundle is comprised of three axial zones each with ten gadolinia rods. These zones are described in the following table and are shown graphically in Figure 2.1.Zone Lattice Geometry Distance from BAF U235 wt% Gadolinia wt%3 AlOT 126" to 149" 4.47 3.5 2 AlOT 96" to 126" 4.57 6.0 1 A1OB 0" to 96" 4.57 6.0 6.1 Geometry Model The ATRIUM-10 fuel assembly parameters are given in Table 4.1. The key fuel pool storage rack parameters are given in Table 4.2. The main KENO storage rack geometry model used for analysis is an infinite array of stainless steel fuel storage boxes with a chevron shaped neutron absorbing insert in each accessible box. All inserts will have the same orientation throughout the entire spent fuel pool; therefore, the fuel assemblies loaded on 2 sides of the perimeter will not be completely enclosed by the inserts (see Figure 4.3). All accessible storage rack cells are modeled with an ATRIUM-10 fuel assembly.6.2 Definition of REBOL Lattices The CASMO-4 lattice depletion calculations are performed at hot operating, uncontrolled,] void history conditionst. The calculation results are based upon the nominal fuel design parameters (defined in Table 4.1) and assume a standard 100 mil fuel channel. Cold xenon-free restart calculations are performed as a function of exposure and void history to establish the highest in-rack reactivity (k.) at any time throughout the life of the fuel lattice. The maximum CASMO-4 in-rack k., of the reference bounding lattices are 0.8843, 0.8869, and* It is demonstrated in Appendix B that the ATRIUM-10 reference design in the spent fuel pool geometry without Boraflex and with NETCO-SNAP-IN inserts is more reactive than the other fuel types used in the LaSalle reactors.t [].AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbina Inserts and Without Boraflex ANP-2843(NP) Revision 1 Paae 6-2 0.9185, for Zones 1 though 3 respectively. These limiting results are based upon a water temperature of 4 °C, 40% void history, and lattice exposures of 16.5, 16.0, and 11.5 GWd/MTU, respectively for each axial zone. The results of the CASMO-4 comparison calculations are summarized in Table 6.1.The following table is provided to summarize the differences between the fuel assembly and lattice names used in this evaluation. Fuel Lattice Type Description ATRIUM-10 REBOL Lattices Defined for use in the KENO calculations, 2.66 wt% U235 (Zone 1), 2.72 wt% (Zone (top and bottom zone geometries) 2), and 3.05 wt% U235 (Zone 3 ), no gadolinia, uniform enrichment distribution, selected to be at least 0.01 Ak more reactive than the reference bounding lattices.ATRIUM-10 Reference Bounding Lattices The most reactive lattices supported by this evaluation with distributed enrichment (top and bottom zone geometries) distribution, 4.57 wt% U235 with 10 Gd 2 03 rods at 6.0 wt% gadolinia (Zones 1 and 2), and 4.47 wt% U235 with 10 Gd 2 0 3 rods at 3.5 wt% gadolinia in Zone 3. These lattices are defined to establish the minimum reactivity required for the REBOL lattices.As-Fabricated Assemblies The actual assemblies built for and/or used in the LaSalle reactors. CASMO-4 in-rack (ATRIUM-10, ATRIUM-9, GE14, and GE 8x8) k. comparisons are included in Appendix B.In support of the KENO rack calculations, reactivity equivalent beginning of life (REBOL) lattice enrichments are selected using the top and bottom ATRIUM-10 lattice geometries. Two REBOL lattices are created with the ATRIUM-10 top geometry and one with the ATRIUM-10 bottom geometry. The REBOL lattices have the same enrichment in all rods and no gadolinia. The REBOL lattice enrichments as well as the CASMO-4 in-rack k. at 4 0 C are shown in Table 6.1.As discussed in the methodology section, a 0.01 Ak adder is included in the generation of the REBOL lattices to address CASMO-4 code, geometry, material, and depletion uncertainties. AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page 6-3 6.3 Storage Array Reactivity For the general KENO rack array calculations, an infinite array of fuel storage cells was assumed -using periodic boundary conditions in all three directions. All fuel locations in the rack array model contain an ATRIUM-10 REBOL assembly comprised of a 3.05 wt% U-235 top zone (above 126"), a 2.72 wt% U-235 intermediate zone (96" to 126"), and a 2.66 wt% U-235 bottom zone (below 96"). The array k-eff is highest when the assembly is centered in the available water space in the storage cell and the assembly orientation shown in Figure 4.2 is as limiting as the other 3 simple rotation possibilities. Calculations were performed at temperatures of 4 'C, 20 'C, 100 'C and 120 °C*. As shown in Table 6.2, the limiting base case KENO k-eff is 0.916.The KENO model assumes a standard 100 mil fuel channel. The array k-eff is about 0.006 Ak lower when the fuel channels are removed.t There is no significant difference in array reactivity between the AREVA standard 100 mil fuel channel and the AREVA advanced fuel channel.*As discussed in Section 4.2 and illustrated in Figure 4.3, assemblies loaded in storage cells on the top and left hand sides of the figure will not be completely surrounded by neutron absorbing inserts. (The entire spent fuel pool is shown in Figure 1.1 of Reference 1 and contains irregular regions). Since the main KENO calculations used an infinite 3-D model it is necessary to evaluate whether the lack of neutron absorbing inserts on these 2 edges of the pool will have a non-conservative effect. This was evaluated using finite 3-D KENO calculations with a 24x24 array of storage cells surrounded by water and concrete, (each cell contained an assemblya and' .a NETCO-SNAP-IN insert). The initial case modeled the condition where all fuel assemblies are enclosed by inserts and was achieved by adding additional inserts§ along the top and left hand edges of the array in the outer water region. The comparison case modeled the more realistic condition where the additional inserts in the water region were removed. Based on this comparison the infinite lattice results will be increased by 0.001 Ak to account for this peripheral edge condition and to ensure conservative results are reported.* 120 0 C addresses the higher temperature conditions that are possible with fuel assemblies near the bottom of a 30 to 40 foot pool of water.t This is because the storage array is over-moderated between the fuel assemblies.

This analysis also supports the use of a standard 80 mil fuel channel.§ The additional inserts were modeled outside of the storage rack array with the same overall spacing and orientation as the inserts in the storage rack cells.AREVA NP Inc.

AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel ANP-2843(NP) Storage Pool Criticality Safety Analysis with Revision 1 Neutron Absorbing Inserts and Without Boraflex Page 6-4 The limiting conditions for the KENO rack calculations are shown in Table 6.2. Except as specifically noted, the reactivity values presented in Tables 6.1 and 6.2 do not include adjustments for uncertainties or code biases. Section 6.6 presents the determination of the upper limit 95/95 reactivity for the storage rack array.6.4 Uncertainties Uncertainties associated with defining bounding REBOL lattices are addressed in Appendix D.Specifically uncertainties associated with CASMO-4 code depletion and modeling capabilities are included within the REBOL definition process.The unadjusted reactivity result reported in Table 6.2 is based upon the nominal bundle position and orientation in the storage rack shown in Figure 4.2. Simple rotation of the assembly or movement within the storage cell does not produce higher (statistically significant) results. As discussed in Section 6.3 a 0.001 Ak adder has been identified to account for the lack of B-10 absorber along 2 peripheral edges of the storage rack array. The manufacturing tolerance values and the calculated reactivity uncertainties for the ATRIUM-1 0 fuel are shown in Table 6.3. The gadolinia manufacturing uncertainty effect on reactivity was evaluated with a combination of KENO V.a and CASMO-4. All other uncertainties reported in Table 6.3 were evaluated with KENO V.a. The ATRIUM-10 rack calculations are conservatively performed for a minimum B1 0 areal density of the insert. BOL dimensions have been assumed, except the fuel rod pitch and channel bulge results are based upon conservative spacer and channel growth dimensions.*

6.5 Abnormal

and Accident Conditions In addition to the nominal storage cell arrangement, abnormal and accident conditions have also been considered. All Ak values provided in this section are based upon comparative KENO V.a calculations -only the most limiting scenario will be reflected in the k 9 5/9 5 calculation in Section 6.6.For the misloaded assembly scenario, only the misplacement of a fuel assembly outside of and adjacent to the storage rack was analyzed because spent fuel pool rack drawings show that there is no gap between the racks wide enough to allow insertion of an assembly. No fuel* The presence of activated corrosion and wear products (CRUD) is neglected because most of these compounds have higher neutron absorption cross sections than water.AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel ANP-2843(NP) Storage Pool Criticality Safety Analysis with Revision 1 Neutron Absorbing Inserts and Without Boraflex Page 6-5 channel was present on the misplaced assembly* and it was placed up against the stainless steel storage rack wall in a location where there is no neutron absorbing insert (see Figure 4.3)between the misplaced assembly and the adjacent assembly. Because this occurs on the edge of the rack array, where neutron leakage is high, only a small reactivity increase was observed (less than 0.001 Ak).The situation where a single neutron absorbing insert is missing from an interior position of the storage rack was also evaluated. This was found to be the most reactive accident condition with a worth of 0.003 Ak.The positioning of the assemblies within the storage cell was also evaluated for conditions with and withoutt a fuel channel. (This bounds the likely condition of an assembly being centered at the bottom and leaning against the storage cell wall at the top). Different configurations that pushed the assemblies toward each other in several combinations were investigated. The most reactive condition was found to occur when all assemblies are centered in the water region of the storage cell with a fuel channel installed. Since this is the nominal condition assumed for this analysis the effect of abnormal (or eccentric) assembly positioning is zero.The orientation of the bundles within the storage rack is not restricted; therefore, the slightly asymmetric nature of the ATRIUM-10 fuel lattices has the potential to increase the pool reactivity if an optimal configuration is achieved. The 4 simple uniform rotation conditions were considered in Section 6.3, and 5 more complicated rotational combinations were evaluated as abnormal conditions. These complicated combinations investigated the effects of how rows, columns, and groups of assemblies could be oriented. From these cases, the worth of abnormal assembly orientation was found to be less than 0.001 Ak. This value is from a case where four rotation conditions are combined.For the case of dropping a fuel assembly onto an assembly in the storage rack, the deformation of either assembly will not be sufficiently large to exceed the reactivity worth of these other limiting accident conditions. This is because it only involves 2 assemblies in a localized area.There will also be no effect on the array reactivity when the dropped assembly comes to rest in a horizontal or inclined position on top of the storage rack because the dropped assembly will* To also evaluate minimum separation scenarios. t To also evaluate minimum separation scenarios. AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel ANP-2843(NP) Storage Pool Criticality Safety Analysis with Revision 1 Neutron Absorbing Inserts and Without Boraflex Page 6-6 be neutronically isolated from the fuel in the storage cells (greater than 12 inches of water between the dropped assembly and the top of the active fuel zone of the fuel in the storage rack).A dropped assembly falling into an empty storage cell would potentially deform the baseplate at the bottom of the storage cell. This could place the dropped assembly at a lower elevation than the other assemblies in the array which would decrease the array reactivity because of increased neutron leakage. If the impact deformed the dropped assembly a higher reactivity condition could be achieved; however, it would be bounded by these other limiting accident conditions because it is limited to a localized area.6.6 Determination of Maximum Rack Assembly k-eff For the ATRIUM-1 0 fuel design with REBOL lattice enrichments of 3.05 wt% U-235 (above 126"), 2.72 wt% U-235 (from 96" to 126"), and 2.66 wt% U-235 (from 0" to 96"), the maximum KENO calculated in-rack reactivity from Table 6.2 is 0.916. This k-eff value is used with the following equation to determine the upper limit 95/95 reactivity: k 9 5/9 5 = keff + biasm + Aksys + (C 2 0Gk 2 + Cm 2 0"m 2 + C 2 0"sys 2 + Akto1 2)1/2, where: keff = in-rack reactivity from KENO V.a, (0.916, Table 6.2)biasm = KENO V.a validation methodology bias (0.00542, page C-18)Aksys = Summation of applicable system variables: maximum k-eff increase due to abnormal and accident conditions from Section 6.5 (0.003) and edge effect adder from Section 6.3 (0.001).C = 95% confidence level consistent with KENO V.a (2)C M = 95/95 one-sided tolerance multiplier for a sample size of 100 (1.927)-= k-eff standard deviation from KENO V.a, (0.001, Table 6.2)CrM = KENO V.a methodology uncertainty (0.00511, page C-18)c'sys = ('sysl 2 + Gsys22 ... + Osys-n 2)1/2%, for Aksys uncertainties Akto, = Statistical combination of manufacturing reactivity uncertainties (0.0105,Table 6.3)** The uncertainty value for non-ATRIUM-10 fuel types will not differ significantly. AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel ANP-2843(NP) Storage Pool Criticality Safety Analysis with Revision 1 Neutron Absorbing Inserts and Without Boraflex Page 6-7 The following table provides a summary of the Aksy, and uoy, parameters applicable to this analysis. (The o values are standard deviation results from KENO)Description Aksys Osys Edge Effect (Insert Orientation, Section 6.3) 0.001 0.0007 Limiting Accident (Missing Insert, Section 6.5) 0.003 0.0006 Combined Values 0.004 0.0009 The standard deviations and tolerance uncertainties are included as the square root of the sum of the squares since they represent independent events. Solving for k 9 5, 9 5 yields a 95/95 upper limit k-eff of 0.940. The above determination of the upper limit 95/95 k-eff is consistent with the method documented in Reference 8 and allows one to state that at least 95% of the normal population is less than the 95/95 k-eff value calculated with a 95% confidence. The results demonstrate the postulated configuration with the ATRIUM-10 REBOL assembly lattices meets the NRC criticality safety acceptance criterion that the array k-eff under the worst credible conditions is < 0.95. Since the REBOL infinite lattices have a higher reactivity than the reference bounding lattices as shown in Table 6.1, the reference bounding lattices also meet the k-eff < 0.95 regulatory limit.6.7 Uniform vs. Distributed Enrichment Distributions A uniform enrichment distribution increases the BWR lattice reactivity because low enriched rods in the corners of the lattice are replaced with rods at an average enrichment level. Relative to the reference bounding lattices described in Table 6.1 a uniform enrichment distribution is more reactive by 0.005 to 0.007 Ak. This increase in reactivity is primarily due to increasing the enrichment in corner pins. This does not affect the results of this evaluation since a BWR assembly will always require low enrichments in the corners to maintain margin to LHGR and CPR limits.6.8 Arrays of Mixed BWR Fuel Types It is shown in Table B.1 that the ATRIUM-10 reference bounding lattices are equal to or more reactive in the in-rack configuration than the limiting lattices of the legacy fuel. Because the AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel ANP-2843(NP) Storage Pool Criticality Safety Analysis with Revision 1 Neutron Absorbing Inserts and Without Boraflex Page 6-8 GEl4, ATRIUM-9, and ATRIUM-10 lattices have similar water and fuel characteristics the neutron energy spectra will be similar for these lattice types. Additionally, it is also shown in Table B.6 that the legacy 8x8 lattices have margin relative to the limiting lattices. It then follows that the ATRIUM-10 lattices used in this evaluation can reasonably represent past assembly fuel types.The assembly enrichment and gadolinia limitations defined in Table 2.1 will be applied to all future ATRIUM-10 fuel assemblies that are built for LaSalle Unit 1 and Unit 2. Therefore, there will not be a more reactive assembly to consider in a misloaded assembly accident and an array composed of a mixture of these fuel types will not exceed the reactivity calculated for an array of limiting ATRIUM-10 assemblies.

6.9 Inaccessible

Storage Locations There are fuel storage locations around the edges of the LaSalle Unit 2 spent fuel pool which are physically inaccessible primarily due to crane interference with piping above the fuel storage racks. These locations will not contain an insert or a fuel assembly. The impact on the storage array k-eff was evaluated for different geometric configurations of empty storage locations without inserts.The evaluation was performed using a 24X24 storage configuration. Originally all storage locations were fully loaded and contained inserts. Additional evaluations were completed with various storage locations containing neither fuel assemblies nor inserts. The locations and configurations evaluated are given in Table 6.4. These locations were selected to represent the irregular edge shape of the storage pool as well as configurations which could occur during the process of installing the inserts. For all cases the fully loaded array with inserts had the highest k-eff. The array reactivity is lower (by up to 0.002 Ak) with no neutron absorbing inserts and no fuel assemblies as defined by the geometries in Table 6.4. Therefore, empty cell locations without an assembly and without an insert do not increase the storage array k-eff.6.10 Interfaces between Areas with Different Storage Conditions As the inserts are installed the storage pool will become a mixture of degraded Boraflex regions and insert regions. The criticality safety evaluations for each of these loading configurations has demonstrated that on an independent (or single region) basis the storage pool multiplication factor is less than the 0.95 regulatory limit. The multiplication factor for a mixture of these AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page 6-9 regions would be expected to also remain below 0.95 if the net transfer of neutrons from one region to another does not increase significantly. Exelon commits to expand the placement of inserts into one row and one column of the adjacent region as necessary to completely surround all assemblies that are part of the insert region with four wings of the NETCO-SNAP-IN inserts*. As addressed in Section 6.8, the reactivity of future ATRIUM-10 fuel assemblies will not exceed the reference bounding assembly of this analysis. With these restrictions in place, the system k-eff of a pool comprised of insert regions mixed with degraded Boraflex regions will be lower than the maximum reported single region value. This occurs because replacement of a large portion of the storage area with another that has a lower multiplication factor decreases the multiplication factor of the entire storage area. KENO evaluations have demonstrated that the resulting k-eff for a system composed of two regions is between that of the individual systems composed of single regions.The overall conclusion from this multi-region analysis is that the spent fuel pool will have a k95/95 value less than or equal to 0.95. This conclusion is reached without crediting residual boron within the insert region.* An exception to this would be peripheral regions of the rack that have no adjacent region.AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page 6-10 Table 6.1 Summary of CASMO-4 Maximum Reactivity Results for the ATRIUM-10 Fuel Assembly Characteristics of the Reference Bounding Fuel Lattices ATRIUM-10 lattice 4.57 wt% U-235 distributed enrichment up to 126" 4.47 wt% U-235 distributed enrichment above 126" 10 gadolinia rods with 3.5 wt% Gd 2 0 3 above 126" and 6.0 wt% Gd 2 0 3 from 0" to 126" Standard 100 mil Channel No xenon in cold calculations Top and bottom lattice geometry explicitly modeled Reflective boundary for in-core Periodic boundary for in-rack Limitinq Conditions Top Lattice Exposure 11.5 GWd/MTU 40% void history Intermediate Lattice Exposure 16.0 GWd/MTU 40% void history Bottom Lattice Exposure 16.5 GWd/MTU 40% void history Calculated Boundina Lattice Reactivitv Condition In-Core, 20 0 C (68°F)In-Rack*, 201C (68°F)In-Rack, 41C (39.2°F)Top Lattice 1.288 Maximum k Intermediate Lattice 1.244 0.917 0.9185 0.886 0.8869 Bottom Lattice 1.241 0.883 0.8843* In-Rack implies the Unit 2 spent fuel pool without Boraflex and with NETCO-SNAP-IN inserts.AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page 6-11 Table 6.1 Summary of CASMO-4 Maximum Reactivity Results for the ATRIUM-10 Fuel Assembly (Continued) REBOL Lattice Conditions ATRIUM-10 top or bottom geometry with uniform enrichment distribution 3.05 wt% U-235 (above 126")2.72 wt% U-235 (from 96" to 126")2.66 wt% U-235 (from 0" to 96")No gadolinia BOL (zero exposure)Standard 100 mil Channel No xenon Top and bottom lattice geometry explicitly modeled Reflective boundary for in-core Periodic boundary for in-rack Calculated REBOL Lattice Reactivity Condition In-Core, 20 0 C (68-F)In-Rack*, 201C (68°F)In-Rack, 4 0 C (39.2 0 F)Top Lattice 1.342 0.926 Maximum k Intermediate Lattice 1.309 0.895 Bottom Lattice 1.308 0.892 0.929 0.898 0.895* In-Rack implies the Unit 2 spent fuel pool without Boraflex and with NETCO-SNAP-IN inserts.AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbina Inserts and Without Boraflex ANP-2843(NP) Revision 1 Paae 6-12 Table 6.2 Summary of KENO V.a Maximum In-Rack Reactivity for ATRIUM-10 Fuel Fuel Assembly ATRIUM-10 top geometry REBOL Lattice (above 126")3.05 wt% U-235 uniform enrichment ATRIUM-10 top geometry REBOL Lattice (96" to 126")2.72 wt% U-235 uniform enrichment ATRIUM-10 bottom geometry REBOL Lattice (from 0" to 96")2.66 wt% U-235 uniform enrichment No gadolinia No xenon Zero exposure Standard 100 mil Channel*Top and bottom lattice geometry explicitly modeled Periodic boundary conditions Storage Array Configuration 13x13 array with periodic boundary conditions in all directions Storage cell pitch preserved across storage rack boundaries Neutron absorbing, chevron shaped insert in each storage cell Assembly centered in cell water volume (not centered relative to stainless steel box)4 0 C moderator and fuel temperatures Maximum Rack Reactivity Description k-eff In-Rack 4°C (39.2'F) k-eff 0.916 +/- 0.001 Maximum k 9 59 5 Reactivity (including uncertainties, biases, manufacturing tolerances and worst accident or abnormal 0.940 loading conditions)

Relative to array reactivity there is no significant difference between the 100 mil and the AREVA Advanced Fuel Channel.AREVA NP Inc.

AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbina Inserts and Without Boraflex ANP-2843(NP) Revision 1 Paae 6-13 Table 6.3 Manufacturing Reactivity Uncertainties (Based upon BOL conditions using KENO V.a except as noted. Ak results of 0.0007 indicate cases where the differences were less than the uncertainty of the calculation) Quantity Nominal Value Tolerance Aký(Reactivity Uncertainty of Fuel Assembly Tolerance Values)Fuel rod pitch Fuel enrichment Fuel density Channel bulge Pellet diameter Clad diameter -outer/inner Pellet void volumet Gadolinia concentration§ 0.510 in.4.57 wt% U235 96.26% TD 0 0.3413 in.0.3957/0.3480 in.1.2%3.5 wt%6.0 wt%[[[[[I[ I[[I I (Reactivity Uncertainty of Rack Tolerance Values)Areal B-10 density Insert thickness SS wall thickness Storage cell pitch Storage cell inside dimension Statistical combination of uncertaintiestt Reported Value>0.0086 g B1O/cm 2 0.065 in.0.090 in.6.255 in.6.0 in.Min value was used+0.005 in.+0.009 in.+0.020 in.0 I 1 1[ I 0.01 05* Value is based upon component measurements at approximate peak reactivity exposures. t This value is equally valid for a fuel density of 95.85% TD.* This is an insignificant parameter; its effect was combined with the U235 enrichment result.§ The gadolinia uncertainty Ak includes a CASMO-4 based 0.002 Ak adder which accounts for differences at peak reactivity conditions. Calculations confirmed that the storage vault reactivity is not affected by the thickness of the insert.This is expected because the B-10 density is defined as an areal density.tl This is based upon the square root of the sum of the squares for all independent tolerance conditions. AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page 6-14 Table 6.4 Evaluation for Inaccessible Storage Locations Storage Cell Configuration* (X,Y) Location within 24X24 arrayt lx1 Center of array 2x2 Center of array 1Xi NE corner of array 4x1 NE corner of array 2x2 NE corner of array 1X4 NE corner of array 1 X2 Center East side of array 2X1 Center East side of array 2x2 Center East side of array 3x3 Center East side of array 2x2 Center West side of array 1X4 SW corner of array 4X1 SW corner of array* These locations do not contain a neutron absorbing insert or a fuel assembly.t Locations (N, S ,E, or W) are relative to the computer model only.AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel ANP-2843(NP) Storage Pool Criticality Safety Analysis with .... Revision 1 Neutron Absorbing Inserts and Without Boraflex Page 7-1 7.0 Conclusions This analysis demonstrates that all fuel assemblies delivered to the LaSalle Station (both Units 1 and 2) as of July 2009 can be safely stored in the LaSalle Unit 2 spent fuel pool with NETCO-SNAP-IN inserts. Future ATRIUM-10 fuel designs that meet the design requirements specified in Table 2.1 or that can be shown to be bounded by the reference bounding assembly can be safely stored in the LaSalle Unit 2 spent fuel pool. The array k-eff determined herein for the reference assembly, including all uncertainties, biases, manufacturing tolerances and worst accident or abnormal loading conditions is 0.940.AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel ANP-2843(NP) Storage Pool Criticality Safety Analysis with Revision 1 Neutron Absorbing Inserts and Without Boraflex Page 8-1 8.0 References

1. Commonwealth Edison LaSalle Station Unit 2 Spent Fuel Storage Capacity Modification Safety Analysis Report, 8601-00-0084, Revision 8, August 1986.2. NUREG/CR-0200 Revision 6, SCALE Version 4.4 A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation, Oak Ridge National Laboratory, May 2000.3. NUREG/CR-6698, Guide for Validation of Nuclear Criticality Safety Calculational Methodology, Nuclear Regulatory Commission, January 2001.4. US Tool and Die Drawing 8601-7 Revision 4, "Commonwealth Edison Co. LaSalle County Station Unit-2 Spent Fuel Storage Racks Fuel Box Assembly & Groups", released by Sargent & Lundy, April 1990.5. NET-259-03 Revision 5, "Material Qualification of ALCAN Composite for Spent Fuel Storage," Northeast Technology Corp., July 2009.6. Code of Federal Regulations, Title 10, Part 50, Section 68, "Criticality Accident Requirements." 7. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Section 9.1.1 Revision 3 (Criticality Safety of Fresh and Spent Fuel Storage and Handling), U.S. Nuclear Regulatory Commission, March 2007.8. Design Requirements for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants, ANSI/ANS American National Standard 57.2-1983, American Nuclear Society, October 1983, (withdrawn 1993).9. Criticality Safety Criteria for the Handling, Storage and Transportation of LWR Fuel Outside Reactors, ANSI/ANS American National Standard 8.17-1984, American Nuclear Society, January 1984, (withdrawn 2004).10. Letter, Brian K. Grimes, Assistant Director for Engineering and Projects Division of Operating Reactors, U.S. Nuclear Regulatory Commission, to All Power Reactor Licensees, "OT Position for the Review and Acceptance of Spent Fuel Storage and Handling Applications," April 14, 1978, as amended by letter, January 18, 1979.11. Letter, Laurence Kopp (Reactor Systems Branch, NRC) to Timothy Collins, Chief (Reactor Systems Branch-NRC),

Subject:

"Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," August 1'9, 1998.12. EMF-2158(P)(A)

Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2, Siemens Power Corporation, October 1999.AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel ANP-2843(NP) Storage Pool Criticality Safety Analysis with Revision 1 Neutron Absorbing Inserts and Without Boraflex Page A-1 Appendix A Sample CASMO-4 Input Tables A.1, A.2, and A.3 provide the in-rack CASMO-4 models for the reference bounding lattices defined by this analysis.ATRIUM-10 fuel which does not conform to the enrichment and gadolinia requirements described in Table 2.1 and Figure 2.1 can be analyzed for storage in the spent fuel pool racks by adapting the CASMO-4 sample inputs presented in Table A.1, A.2 or A.3. For bottom lattices the evaluation should be completed with both [ ] depletions. Intermediate and top lattices should be evaluated at both [ ] depletions. If the lifetime maximum in-rack k. of the new lattice is less than the k. of the corresponding reference bounding lattice, the ATRIUM-10 fuel assembly can be safely stored in the LaSalle Unit 2 Nuclear Power Station spent fuel storage rack.If a different version of CASMO-4 is used, it is recommended that the sample cases for the reference bounding lattices (provided in Tables A.1 through A.3) be re-evaluated to establish that the version of CASMO-4 and the underlying libraries being used are consistent with those used in this report. Small changes, less than 0.005 Ak from the results in this report, are acceptable and can be used to establish new k- limits for comparison to the new lattices (i.e. the comparison should be performed based upon the same calculational basis). Larger changes from the results contained in this report represent more significant changes in the underlying model and may require additional CASMO-4 to KENO benchmarking. AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel ANP-2843(NP) Storage Pool Criticality Safety Analysis with Revision 1 Neutron Absorbing Inserts and Without Boraflex Page A-2 Table A.1 CASMO-4 Input for ATRIUM-10 Top Reference Bounding Lattice TTL

AlOT-4470L-10G35_BL

-.40 VB TFU= 814.3 TMO= 560.3 VOI=40 FUE, 1,10.42349/ 2.5000 FUE, 2,10.42349/ 3.4000 FUE, 3,10.42349/ 4.2000 FUE, 4,10.29433/ 4.4100,64016= 3.5000 FUE, 5,10.42349/ 4.6900 FUE, 6,10.42349/ 4.8000 FUE, 7,10.42349/ 4.9500 BWR, 10,1.29540,13.40612,0.25400,0.66294,0.66294,1.2700,1 THE,0 FUM, 0,2 PIN, 1,0.43345,0.44196,0.50254 PIN, 2,1.67767,1.75006/'MOD','BOX'//-9 PIN, 3,0.44196,0.50254/'COO','COO' LPI 1 13 111 1111 13112 111122 1111222 11111111 131113113 1 1 1 1 1 1 1 1 1 1 LFU 1 2 0 3 7 7 5 4 7 7 5 0 7 7 0 5 4 7 7 0 0 5 7 7 7 0 0 0 3 7 7 7 4 6 3 3 2 0 7 4 7 0 4 7 0 1 2 3 6 5 5 6 3 2 1 PDE, 51.9538, 'KWL'DEP 0.0,0.1,0.5,1,1.5,2,2.5,3,3.5,4,4.5,5,5.5,6,6.5,7,7.5,8,8.5,9,9.5,10,10.5, 11,11.5,12,12.5,13,13.5,14,14.5,15 STA TTL *+LaSalle Rack at 4 deg. C (No BF with Boral Insert)RES, ,0, 9,-15 VOI,00 TMO, 277.1 TFU, 277.1 PDE,0 CNU,'FUE',54135,1.OE-14 BCO 'PER'GAP 4*0.49784 MII 0.05209/5010=100.0 M12 5.8408/347=94.89 1001=0.57 8000=4.54 FST 4*0.16510/4*0.32385/2*'MOD' 5*'MII' 'MOD'/8*'MI2'/ STA END AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel ANP-2843(NP) Storage Pool Criticality Safety Analysis with Revision 1 Neutron Absorbing Inserts and Without Boraflex Page A-3 Table A.2 CASMO-4 Input for ATRIUM-10 Intermediate Reference Bounding Lattice TTL

AIOT-4570L-10G60 BL -.40 VB TFU= 814.3 TMO= 560.3 VOI=40 FUE, 1,10.42349/

2.5000 FUE, 2,10.42349/ 3.6000 FUE, 3,10.42349/ 4.4000 FUE, 4,10.20471/ 4.5500,64016= 6.0000 FUE, 5,10.42349/ 4.8000 FUE, 6,10.42349/ 4.9500 BWR, 10,1.29540,13.40612,0.25400,0.66294,0.66294,1.2700,1 THE, 0 FUM, 0,2 PIN, 1,0.43345,0.44196,0.50254 PIN, 2,1.67767,1.75006/'MOD','BOX'//-9 PIN, 3,0.44196,0.50254/'COO','COO' LPI 1 13 111 1111 13112 111122 1111222 11113113 1 1 1 1 1 1 1 1 1 1 LFU 1 2 0 3 6 6 6 4 6 6 6 0 6 6 0 6 4 6 6 0 0 6 6 6 6 0 0 0 3 6 6 6 4 5 3 3 2 0 6 4 6 0 4 6 0 1 2 3 5 6 6 5 3 2 1 PDE, 51.9538, 'KWL'DEP 0.0,0.1,0.5,1,1.5,2,2.5,3,3.5,4,4.5,5,5.5,6,6.5,7,7.5,8,8.5,9,9.5,10,10.5, 11,11.5,12,12.5,13,13.5,14,14.5,15,15.5,16,16.5,17,17.5,18,18.5,19,19.5,20, 20.5,21,21.5,22,22.5,23,23.5,24,24.5,25 STA TTL *+LaSalle Rack at 4 deg. C (No BF with Boral Insert)RES,,0,11,-25 VOI,00 TMO, 277.1 TFU, 277.1 PDE,0 CNU,'FUE',54135,1.OE-14 BCO 'PER'GAP 4*0.49784 MIl 0.05209/5010=100.0 M12 5.8408/347=94.89 1001=0.57 8000=4.54 FST 4*0.16510/4*0.32385/2*'MOD' 5*'MII' 'MOD'/8*'MI2'/ STA END AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel ANP-2843(NP) Storage Pool Criticality Safety Analysis with Revision 1 Neutron Absorbing Inserts and Without Boraflex Page A-4 Table A.3 CASMO-4 Input for ATRIUM-10 Bottom Reference Bounding Lattice TTL

AIOB-4570L-10G60 BL -.40 VB TFU= 791.6 TMO= 560.3 VOI=40 FUE, 1,10.42349/

2.5000 FUE, 2,10.42349/ 3.6000 FUE, 3,10.42349/ 4.4000 FUE, 4,10.20471/ 4.4600,64016= 6.0000 FUE, 5,10.42349/ 4.8000 FUE, 6,10.42349/ 4.9500 BWR, 10 1.29540,13.40612,0.25400,0.66294,0.66294,1.2700,1 THE,0 FUM, 0,2 PIN, 1,0.43345,0.44196,0.50254 PIN, 2,1.67767,1.75006/'MOD', 'BOX'//-9 LPI 1 11 111 1111 11112 111122 1111222 11111111 1 1 1 1 2 LFU 1 2 3 3 6 6 6 4 6 6 6 6 6 6 0 6 4 6 6 0 0 6 6 6 6 0 0 0 3 6 6 6 4 5 3 3 2 3 6 4 6 6 4 6 3 1 2 3 5 6 6 5 3 2 1 PDE, 51.9538, 'KWL'DEP 0.0,0.1,0.5,1,1.5,2,2.5,3,3.5,4,4.5,5,5.5,6,6.5,7,7.5,8,8.5,9,9.5,10,10.5, 11,11.5,12,12.5,13,13.5,14,14.5,15,15.5,16,16.5,17,17.5,18,18.5,19,19.5,20, 20.5,21,21.5,22,22.5,23,23.5,24,24.5,25 STA TTL *+LaSalle Rack at 4 deg. C (No BF with Boral Insert)RES,,0,11,-25 VOI,00 TMO, 277.1 TFU, 277.1 PDE,0 CNU, 'FUE',54135,1.OE-14 BCO 'PER'GAP 4*0.49784 MIl 0.05209/5010=100.0 M12 5.8408/347=94.89 1001=0.57 8000=4.54 FST 4*0.16510/4*0.32385/2*'MOD' 5*'MII' 'MOD'/8*'MI2'/ STA END AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel ANP-2843(NP) Storage Pool Criticality Safety Analysis with Revision 1 Neutron Absorbing Inserts and Without Boraflex Page B-1 Appendix B Reactivity Comparison for Assemblies Used in the LaSalle Reactors The following tables present a comparison of in-rack CASMO-4 k. values* (without Boraflex and with NETCO-SNAP-IN inserts) of the more reactive lattices of the different fuel assembly types used at or manufactured for the LaSalle Unit 1 or Unit 2 reactors prior to July 2009. For each assembly type, the more reactive lattices have been identified using a comparison of the U235 enrichment levels and the gadolinia concentrations. The comparisons are made based on three axial zones, 0" to 96", 96" to 126", and 126" to 149". The ATRIUM-9 458L-8G6 lattice is the most reactive as-fabricated design from 0" to 96" and from 96" to 126", and the ATRIUM-10T-4444L-12G40 lattice is the most reactive as fabricated design from 126" to 149". In the following tables LSA and LSB refer to LaSalle unit 1 or 2, respectively. The following comparison table shows that the ATRIUM-10 reference bounding lattices described in Table 6.1 are equal to or more reactive than any of the lattices used in the LaSalle reactors. (Also note that the REBOL lattices used in the KENO V.a calculations are more reactive than the reference bounding lattices).

[AREVA NP Inc.

AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page B-2 Table B.1 Lattice Reactivity Comparisons (REBOL, Bounding, and Limiting)Maximum In-Rack k-(CASMO-4)Case Description Lattice Description 4 0 C 20 °C REBOL, Top Lattice 126" to 149" A1OT-305LOG0 0.929 0.926 Reference Bounding Top Lattice 126" A1OT-447L10G35 0.919 0.917 to 149" Limiting As-Fabricated Top Lattice A1OT-4444L12G40 0.907 0.906 126" to 149" REBOL, Intermediate Lattice 96" to A10T-272LOG0 0.898 0.895 126" Reference Bounding Intermediate Al OT-457L1 0G60 0.887 0.886 Lattice 96" to 126" Limiting As-Fabricated Intermediate A9-458L8G6 0.884 0.883 Lattice 96" to 126" REBOL, Bottom Lattice 0" to 96" A1OB-266LOG0 0.895 0.892 Reference Bounding Bottom Lattice 0" A1OB-457L10G60 0.884 0.883 to 96" Limiting As-Fabricated Bottom Lattice A9-458L8G6 0.884 0.883 0" to 96" AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page B-3 Table B.2 ATRIUM-10 Fuel Lattice Reactivity Comparison Maximum In-Rack k. Unit and Case* Latticet (CASMO-4) Cycle 4 0 C 20 0 C 100 OC Loaded T Al0T-4444L12G40 0.907 0.906 0.895 1 A1OT-2111LOGO 0.825 0.822 0.803 LSB CylO 2 A1OT-3947L13G38 0.882 0.881 0.870 LSB Cy13 3 A1OT-4444L12G40 0.907 0.906 0.895 LSA Cyl 3 4 A1OT-4409L10G45 0.907 0.905 0.895 LSB Cy12 4a A1OT-4400L10G45 0.907 0.905 0.895 LSA Cy12 I A9-458L8G6 0.884 0.883 0.875 1 A1OT-2111 LOGO 0.825 0.822 0.803 LSB Cyl 0 5 A1OT-4313L15G65 0.860 0.859 0.850 LSB Cyl0 6 A1OT-4524L13GV70 0.860 0.858 0.849 LSB Cyl3 7 A1OT-4511 L15GV80 0.840 0.839 0.830 LSB Cyl 3 t T, I, and B indicate the most reactive top, intermediate, and bottom lattice cases, respectively. Note that A10T and A10B indicate top and bottom ATRIUM-10 lattice geometry. A9 indicates ATRIUM-9.AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page B-4 Table B.2 ATRIUM-10 Fuel Lattice Reactivity Comparison (Continued) Maximum In-Rack k. Unit and Case* Latticet (CASMO-4) Cycle 4 0 C 20 0C 100 0C Loaded B A9-458L8G6 0.884 0.883 0.875 8 A10B-1831L-0G0 0.785 0.782 0.764 LSB CylO 9 Al0B-4399L12G65 0.871 0.869 0.860 LSA Cyl3 10 A10B-4537L13GV70 0.857 0.856 0.847 LSB Cyl3 11 A1OB-4510L13G75 0.863 0.862 0.853 LSA CylO 12 Al1B-4538L13GV80 0.844 0.843 0.834 LSB Cy13 t T, I, and B indicate the most reactive top, intermediate, and bottom lattice cases, respectively. Note that AMOT and A10B indicate top and bottom ATRIUM-10 lattice geometry. A9 indicates ATRI U M-9.AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page B-5 Table B.3 ATRIUM 1OXM Fuel Lattice Reactivity Comparison Maximum In-Rack k. Unit and Case* Latticet (CASMO-4) Cycle 4 °C 20 °C 100 °C Loaded T Al0T-4444L12G40 0.907 0.906 0.895 ---1 DXMT-4056L12G40 0.880 0.879 0.869 LSB Cyl3t I A9-458L8G6 0.884 0.883 0.875 ---2 DXMT-4176L14GV60 0.852 0.851 0.842 LSB Cy13 B A9-458L8G6 0.884 0.883 0.875 ---2 DXMT-4176L14GV60 0.852 0.851 0.842 LSB Cyl 3 3 DXMB-4365L14GV80 0.840 0.839 0.830 LSB Cyl 3* T, I, and B indicate the most reactive top, intermediate, and bottom lattice cases, respectively. t Note that AlOT and A10B indicate top and bottom ATRIUM-10 lattice geometry. A9 indicates ATRIUM-9.8 ATRIUM 1OXM lead use assemblies have been manufactured as part of the reload fuel for LaSalle Unit 2 Cycle 13.AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page B-6 Table B.4 ATRIUM-9 Fuel Lattice Reactivity Comparison Maximum In-Rack k. Unit and (CASMO-4)Case* Latticet Cycle 4 0 C 20 °C 100 °C Loaded T A10T-4444L12G40 0.907 0.906 0.895 I & B A9-458L8G6 0.884 0.883 0.875 1 A9-396L8G5 0.875 0.874 0.865 LSA&B Cy9 2 A9-458L8G6 0.884 0.883 0.875 LSA&B Cy9 3 A9-459L12G7 0.870 0.869 0.861 LSA Cy9 4 A9-459L12G8 0.858 0.857 0.850 LSA Cy9* T, I, and B indicate the most reactive top, intermediate, and bottom lattice cases, respectively. t Note that AMOT indicates top ATRIUM-10 lattice geometry and A9 indicates ATRIUM-9.AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbina Inserts and Without Boraflex ANP-2843(NP) Revision 1 Paae B-7 Table B.5 GE14 Fuel Lattice Reactivity Comparison Maximum In-Rack k. Unit and (CASMO-4)Case* Latticet Cycle 4 °C 20 °C 100 0C Loaded T A1OT-4444L12G40 0.907 0.906 0.895 1 GE14-429L6G70-9G60 0.849 0.847 0.838 LSB Cyl 1 2 GE14-430L2G80-7G70-5G60 0.844 0.843 0.834 LSB Cyl 1 3 GE14-446L-10G80-4G70 0.844 0.842 0.834 LSA Cyl 1 I A9-458L8G6 0.884 0.883 0.875 1 GE14-429L6G70-9G60 0.849 0.847 0.838 LSB Cyl 1 2 GE14-430L2G80-7G70-5G60 0.844 0.843 0.834 LSB Cyl 1 3 GE14-446L-10G80-4G70 0.844 0.842 0.834 LSA Cyl 1 B A9-458L8G6 0.884 0.883 0.875 4 GE14-435L6G70-9G60 0.841 0.840 0.830 LSB Cyl 1 5 GE14-437L2G80-7G70-5G60 0.834 0.832 0.823 LSB Cyl 1 6 GE14-451L10G80-4G70 0.834 0.833 0.824 LSA Cyl 1 6a GE14-451L11G80-4G70 0.842 0.841 0.832 LSA Cyl 1* T, I, and B indicate the most reactive top, intermediate, and bottom lattice cases, respectively. t Note that AlOT indicates top ATRIUM-10 lattice geometry and A9 indicates ATRIUM-9. GE14 indicates GE14 geometry.AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbina Inserts and Without Boraflex ANP-2843(NP) Revision 1 Paae B-8 Table B.6 GE 8x8 Fuel Lattice Reactivity Comparison Maximum In-Rack k. Unit and (CASMO-4)Case* Latticet Cycle 4 °C 20 °C 100 °C Loaded T A1OT-4444L12G40 0.907 0.906 0.895 ---I & B A9-458L8G6 0.884 0.883 0.875 1 8x8_2-319L6G30 0.858 0.857 0.844 LSB Cy3 2 8x8_2-340L7G30 0.869 0.867 0.855 LSB Cy3 3 8x8_4-338L7G30 0.863 0.861 0.850 LSB Cy5 4 8x8_4-388L8G40 0.875 0.874 0.863 LSA Cy8 t T, I, and B indicate the most reactive top, intermediate, and bottom lattice cases, respectively. Note that AMOT indicates top ATRIUM-10 lattice geometry and A9 indicates ATRIUM-9. 8x8_2 implies an 8x8 lattice with 2 water rods and 8x8_4 indicates an 8x8 lattice with a large internal water rod encompassing the area of 4 pin cells, i.e. GE9 fuel.AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel ANP-2843(NP) Storage Pool Criticality Safety Analysis with Revision 1 Neutron Absorbing Inserts and Without Boraflex Page C-1 Appendix C KENO V.a Bias and Bias Uncertainty Evaluation The purpose of the present analysis is to determine the bias of the keff calculated with the SCALE 4.4a computercode for spent fuel pool criticality analysis. A statistical methodology is used to evaluate criticality benchmark experiments that are appropriate for the expected range of parameters. The scope of this report is limited to the validation of the KENO V.a module and CSAS25 driver in the SCALE 4.4a code package for use with the 44 energy group cross-section library 44GROUPNDF5 for spent fuel criticality analyses.This calculation is performed according to the general methodology described in Reference C.2 (NUREG/CR-6698 "Guide for Validation of Nuclear Criticality Safety Calculational Methodology") that is also briefly described in Section C.1. The critical experiments selected to benchmark the computer code system are discussed in Section C.3. The results of the criticality benchmark calculations, the trending analysis, the basis for the statistical technique chosen, the bias, and the bias uncertainty are presented in Sections C.4-C.7. Final results are summarized in Section C.8.C.1 Statistical Method for Determining the Code Bias As presented in Reference C.2 (NUREG/CR-6698), the validation of the criticality code must use a statistical analysis to determine the bias and bias uncertainty in the calculation of keff. The approach involves determining a weighted mean of keff that incorporates the uncertainty from both the measurement (aexp) and the calculation method (acaic). A combined uncertainty can be determined using the Equation 3 from Reference C.2, for each critical experiment: =yt -` (acalc 2 + C'exp 2)1/2 The weighted mean of keff (kf), the variance about mean (s), and the average total uncertainty of the benchmark experiments (2 ) can be calculated using the weighting factor 1 / o-,2 (see Eq.4, 5, and 6 in Reference C.2). The final objective is to determine the square root of the pooled variance, defined as (Eq. 7 from Reference C.2): AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel ANP-2843(NP) Storage Pool Criticality Safety Analysis with Revision 1 Neutron Absorbing Inserts and Without Boraflex Page C-2-._"2 The above value is used as the mean bias uncertainty, where bias is determined by the relation: Bias = cff -1, if kff is less than 1, otherwise Bias = 0 (Eq.8 from Reference C.2)The approach for determining the final statistical uncertainty in the calculational bias relies on the selection of an appropriate statistical treatment. Basically, the same steps and methods suggested in Reference C.2 for determining the upper safety limit (USL) can be applied also for determining the final bias uncertainty. First, the possible trends in bias need to be investigated. Trends are identified through the use of regression fits to the calculated keff results. In many instances, a linear fit is sufficient to determine a trend in bias. Typical parameters used in these trending analyses are enrichment, H/X or a generic spectral parameter such as the energy of the average lethargy causing fission (EALF).Reference C.2 indicates that the use of both weighted or unweighted least squares techniques is an appropriate means for determining the fit of a function. For the present analysis linear regression was used on both weighted and unweighted keff values to determine the existence of a trend in bias. Typical numerical goodness of fit tests were applied afterwards to confirm the validity of the trend.When a relationship between a calculated keff and an independent variable can be determined, a one-sided lower tolerance band may be used to express the bias and its uncertainty (Reference C.2). When no trend is identified, the pool of keff data is tested for normality. If the data is normally distributed, then a technique such as a one-sided tolerance limit is used to determine bias and its uncertainty. If the data is not normally distributed, then a non-parametric analysis method must be used to determine the bias and its uncertainty (Reference C.2).Similar examples of application of these techniques are included in References C.4 and C.5.AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page C-3 C.2 Area of Applicability Required for the Benchmark Experiments BWR spent fuel pools will primarily contain commercial nuclear fuel in uranium oxide pins in a square array. This fuel is characterized by the typical parameter values provided in Table C.1.These typical values were used as primary tools in selecting the benchmark experiments appropriate for determining the code bias.Benchmark calculations have been made on selected critical experiments, chosen, in so far as possible, to bound the range of variables in the spent fuel rack analyses. In rack designs, the most significant parameters affecting criticality are: (1) the fuel enrichment, (2) the 1 0 B loading in the neutron absorber, and (3) the lattice spacing. Other parameters have a smaller effect but have been also included in the analyses.One possible way of representing the data is through a spectral parameter that incorporates influences from the variations in other parameters. Such a parameter is computed by KENO V.a, which prints the "energy of the average lethargy causing fission" (EALF). The expected range for this parameter in the analyses was also included in Table C.1.Table C.1 Range of Values of Key Parameters in Spent Fuel Pool Parameter Fissile material -Physical/Chemical Form Enrichment Moderation/Moderator Lattice Pitch Clad Anticipated Absorber/Materials H/X ratio Reflection Neutron Energy Spectrum (Energy of the Average Lethargy Causing Fission)Ranqe of Values U0 2 rods natural to 5.00 wt% U-235 Heterogeneous/Water Square 1.2 to 1.45 cm Zircaloy Aluminum, Boron Stainless Steel 0 to 473 Water, Stainless Steel 0.1 to 2.5 eV AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel ANP-2843(NP) Storage Pool Criticality Safety Analysis with Revision 1 Neutron Absorbing Inserts and Without Boraflex Page C-4 C.3 Description of the Criticality Experiments Selected The set of criticality benchmark experiments has been constructed to accommodate large variations in the range of parameters of the rack configurations and also to provide adequate statistics for the evaluation of the code bias.One hundred critical configurations were selected from various sources. These benchmarks include configurations performed with lattices of U0 2 fuel rods in water having various enrichments and moderating ratios (H/X). A set of MOX criticality benchmarks is also included in the present set. The area of applicability (AOA) is established within this range of benchmark experiment parameter values.A brief description of the selected benchmark experiments is presented in Table C.2. The table includes the references where detailed descriptions of the experiments are presented. AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Paqe C-5 Table C.2 Descriptions of the Critical Benchmark Experiments Experiment Measure cy exp Brief Description Neutron Absorber Reflector Case Name d keff NUREGICR-0073 PNL experiments (Reference C.3) _______________ ______________ c004 1.0000 0.0020 c005b 1.0000 0.0018 c006b 1.0000 0.0019 c007a 1.0000 0.0021 c008b 1.0000 0.0021 c009b 1.0000 0.0021 cOlOb 1.0000 0.0021 cO11b 1.0000 0.0021 cO12b 1.0000 0.0021 cO13b 1.0000 0.0021 cO14b 1.0000 0.0021 c029b 1.0000 0.0021 c03Ob 1.0000 0.0021 U0 2 pellets with 4.31 wt% Cluster of fuel rods on a 25.4 mm pitch. Moderator; water or borated water.Various separation distances used between clusters. Those so indicated have plates of neutron absorbing material poison placed between clusters of fuel rods.None 0.625 cm Al plates 0.625 cm Al plates 0.302 cm SS-304L plates 0.298 cm SS-304L absorber plates with 1.05 wt % or 1.62 wt% B 0.485 cm SS304L plates Zircaloy-4 absorber plates Boral absorber Water and acrylic plates as well as a biological shield serve as primary reflector material. A minor contribution comes from the channel that supports the rod clusters and the 9.52 mm carbon cthal t~nnl minll c031b 1.0000 0.0021 aclp3 1.0000 0.0006 aclp4 0.9999 0.0006 aclp5 1.0000 0.0007 aclp6 1.0097 0.0012 aclp7 0.9998 0.0009 aclp8 1.0083 0.0012 aclp9 1.0030 0.0009 aclpl0 1.0001 0.0009 aclplla 1.0000 0.0006 aclpllb 1.0007 0.0001 aclpllc 1.0007 0.0006 aclplld 1.0007 0.0006 aclplle 1.0007 0.0006 aclpllf 1.0007 0.0006 aclpllg 1.0007 0.0006 aclp12 1.0000 0.0007 aclp13 1.0000 0.0010 aclp13a 1.0000 0.0010 aclp14 1.0001 0.0010 acIp15 0.9998 0.0016 aclp16 1.0001 0.0019 aclp17 1.0000 0.0010 aclp18 1.0002 0.0011 aclp19 1.0002 0.0010 aclp2O 1.0003 0.0011 r-nrlcnments OT L.143i- WtLIo U 3x3 array of fuel clusters.Various B 4 C pins and stainless steel and boron-aluminum sheets were used as neutron absorbers. Cases so indicated also had dissolved boron in the water moderator. iNone 1037 ppm boron 764 ppm boron None None None None None None 143 ppm boron 510 ppm boron 514 ppm boron 501 ppm boron 493 ppm boron 474 ppm boron 462 ppm boron 432 ppm boron 217 ppm boron 15 ppm boron 28 ppm boron 92 ppm boron 395 ppm boron 121 ppm boron 487 ppm boron 197 ppm boron 634 ppm boron 320 ppm boron 72 ppm boron vvaier ana aluminum base plate are the primary reflective materials in the experiments. Minor contribution from the steel tank walls.aclp21 0.9997 0.0015 AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page C-6_____164_5-4 eie~ t~Rf ic;~~)J rcon0l 1.0007 0.0006 rcon02 1.0007 0.0006 rcon03 1.0007 0.0006 rcon04 1.0007 0.0006 rcon05 1.0007 0.0006 rcon06 1.0007 0.0006 rcon07 1.0007 0.0006 rcon08 1.0007 0.0006 rcon09 1.0007 0.0006 rconl0 1.0007 0.0006 rconl1 1.0007 0.0006 rcon12 1.0007 0.0006 rcon13 1.0007 0.0006 rcon14 1.0007 0.0006 rcon15 1.0007 0.0006 rcon16 1.0007 0.0006 rcon17 1.0007 0.0006 rcon18 1.0007 0.0006 rcon19 1.0007 0.0006 rcon20 1.0007 0.0006 2.46 wt% ...U 5x5 array of fuel cluster. Rod pitch between 1.2093 cm and 1.4097 cm. Cases so indicated also had dissolved boron in the water moderator. 435 ppm boron 426 ppm boron 406 ppm boron 383 ppm boron 354 ppm boron 335 ppm boron 361 ppm boron 121 ppm boron 886 ppm boron 871 ppm boron 852 ppm boron 834 ppm boron 815 ppm boron 781 ppm boron 746 ppm boton 1156 ppm boron 1141 ppm boron 1123 ppm boron 1107 ppm boron 1093 ppm boron 1068 ppm boron 191 nnm hnrnn Water and aluminum base plate are the primary reflective materials in the experiments. Minor contribution from the steel tank walls.rcon2l rrcrn9$A 1.0007 1 nnn7 0.0006 n Wnnn CEA Vaduci0 mdis0l mdis02)ne ie actual;00 0.00 mdis03 1.0000 0.0014 mdis04 1.0000 0.0014 mdis05 1.0000 0.0014 mdis06 1.0000 0.0014 mdis07 1.0000 0.0014 mdis08 1.0000 0.0014 mdis09 1.0000 0.0014 mdisl0 1.0000 0.0014 mdisl 1 1.0000 0.0014 mdis12 1.0000 0.0014 mdis13 1.0000 0.0014 mdis14 1.0000 0.0014 mdis15 1.0000 0.0014 mdis16 1.0000 0.0014 mdis17 1.0000 0.0014 mdis18 1.0000 0.0014 CEA Valduc Critical Mass Laboratory experiments. A key aspect of these experiments was to examine the reactivity effects of differing densities of hydrogenous materials within a cross shaped channel box placed between a two by two array of fuel rod assemblies. reflector boundaries vary from case to case.The assemblies each consisted of an 18 x 18 array of aluminum alloy clad fuel U02 pellet columns.mdis19 1.0000 0.0014 AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbina Inserts and Without Boraflex ANP-2843(NP) Revision 1 Paqe C-7 LEU-CQMP-TERM-022, -024, -025 Experiments (Reference C. 1)leuct022-02 1.0000 0.0046 9.83 and 7.41 wt% enriched U02 None Water is the leuct022-03. 1.0000 0.0036 rods of varying numbers in primary reflector. leuct024-01 1.0000 0.0054 hexagonal and square lattices in Minor leuct024-02 1.0000 0.0040 water. contribution from leuct025-01 1.0000 0.0041 the steel tank leuct025-02 1.0000 0.0044 walls.Mixed Oxide (Reference C.1, Experi ient MlX-COMP-THERM 002)epri70b 1.0009 0.0047 Experiments with mixtures of 687.9 ppm B Reflected by (PNL-31) natural U02-2wt%PuO2 water and Al.epri70un 1.0024 0.0060 (8%240Pu). 1.7 ppm B (PNL-30) Square pitched lattices, with epri87b 1.0024 0.0024 1.778 cm, 2.2098 cm, and 1090.4 ppm B (PNL-33) 2.5146 cm pitch in borated or epri87un 1.0042 0.0031 pure water moderator. 0.9 ppm B (PNL-32)epri99b 1.0029 0.0027 767.2 ppm B (PNL-35)epri99un 1.0038 0.0025 1.6 ppm B (PNL-34)Mixed Oxide (Reference C.1, Experi ent MIX-COMP-Ten1ERM 003).saxtn 104 1.0000 0.0023 Experiments with mixtures of None Reflected by (case 6) natural U02-6.6wt%PuO2 water and Al.saxtn56b 1.0000 0.0054 mixed-oxide (MOX), square- 337 ppm B (case 3) pitched, partial moderator height saxtn792 1.0049 0.0027 lattices. None (case 5) Moderator: borated or pure saxton52 1.0028 0.0072 water moderator. None (case 1)saxton56 1.0019 0.0059 None (case 2)(PNL-35)C.4 Results of Calculations with SCALE 4.4.a The critical experiments described in Section C.3 were modeled with the SCALE 4.4a computer system. The resulting keff and calculational uncertainty, along with the experimental keff and experimental uncertainty are tabulated in Table C.3. The parameters of interest in performing a trending analysis of the bias (Including EALF calculated by SCALE 4.4a) are also included in the table.AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page C-8 Table C.3 SCALE 4.4a Results for the Selected Benchmark Experiments No Case name 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 37 38 39 40 41 42 43 44 45 46 47 c00 4 c005b c006b c007a c008b c009b cOlOb cO1 lb cO12b cO13b cO14b c029b c03Ob c031 b aclp1 aclp2 ac1p3 aclp4 ac1p5 aclp6 aclp7 aclp8 aclp9 aclplO acpl la acpl lb acpllc acpl ld acpl le acpllf acpl lg aclp12 aclpl3 acp13a aclp14 acIp15 aclp16 aclp17 aclp18 aclp19 aclp2O aclp21 rcon0l rcon02 rcon03 rcon04 rcon05 Benchmark values SCALE 4.4a Calculated Values keff Texp keff acalc 1.0000 0.0020 0.9966 0.0008 1.0000 0.0018 0.9950 0.0008 1.0000 0.0019 0.9964 0.0008 1.0000 0.0021 0.9973 0.0009 1.0000 0.0021 0.9966 0.0008 1.0000 0.0021 0.9967 0.0008 1.0000 0.0021 0.9977 0.0008 1.0000 0.0021 0.9949 0.0009 1.0000 0.0021 0.9967 0.0008 1.0000 0.0021 0.9969 0.0008 1.0000 0.0021 0.9958 0.0008 1.0000 0.0021 0.9972 0.0008 1.0000 0.0021 0.9972 0.0009 1.0000 0.0021 0.9993 0.0009 1.0002 0.0005 0.9912 0.0007 1.0001 0.0005 0.9951 0.0006 1.0000 0.0006 0.9958 0.0006 0.9999 0.0006 0.9889 0.0008 1.0000 0.0007 0.9906 0.0007 1.0097 0.0012 0.9899 0.0009 0.9998 0.0009 0.9891 0.0008 1.0083 0.0012 0.9873 0.0007 1.0030 0.0009 0.9908 0.0008 1.0001 0.0009 0.9916 0.0007 1.0000 0.0006 0.9948 0.0007 1.0007 0.0007 0.9947 0.0007 1.0007 0.0006 0.9944 0.0006 1.0007 0.0006 0.9952 0.0007 1.0007 0.0006 0.9940 0.0006 1.0007 0.0006 0.9932 0.0007 1.0007 0.0006 0.9954 0.0007 1.0000 0.0007 0.9930 0.0008 1.0000 0.0010 0.9933 0.0008 1.0000 0.0010 0.9902 0.0007 1.0001 0.0010 0.9891 0.0008 0.9998 0.0016 0.9855 0.0007 1.0001 0.0019 0.9856 0.0007 1.0000 0.0010 0.9899 0.0006 1.0002 0.0011 0.9886 0.0008 1.0002 0.0010 0.9912 0.0006 1.0003 0.0011 0.9899 0.0007 0.9997 0.0015 0.9883 0.0008 1.0007 0.0006 0.9997 0.0007 1.0007 0.0006 1.0004 0.0007 1.0007 0.0006 0.9985 0.0008 1.0007 0.0006 0.9983 0.0007 1.0007 0.0006 1.0002 0.0007 EALF Enr (eV) wt%2 3 5 U 0.1126 4.31 0.1128 4.31 0.1130 4.31 0.1128 4.31 0.1135 4.31 0.1136 4.31 0.1142 4.31 0.1143 4.31 0.1148 4.31 0.1130 4.31 0.1133 4.31 0.1126 4.31 0.1132 4.31 0.1144 4.31 0.1725 2.46 0.2504 2.46 0.1963 2.46 0.1912 2.46 0.1660 2.46 0.1712 2.46 0.1496 2.46 0.1537 2.46 0.1409 2.46 0.1495 2.46 0.1996 2.46 0.1994 2.46 0.2019 2.46 0.2028 2.46 0.2037 2.46 0.2050 2.46 0.2045 2.46 0.1700 2.46 0.1965 2.46 0.1981 2.46 0.2011 2.46 0.2063 2.46 0.1730 2.46 0.2053 2.46 0.1725 2.46 0.2061 2.46 0.1730 2.46 0.1532 2.46 2.4282 2.46 2.4360 2.46 2.4972 2.46 2.4989 2.46 2.4988 2.46 B (ppm)0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1037 764 0 0 0 0 0 0 143 510 514 501 493 474 462 432 217 15 28 92 395 121 487 197 634 320 72 435 426 406 383 354 H/X 255.92 255.92 255.92 255.92 255.92 255.92 255.92 255.92 255.92 255.92 255.92 255.92 255.92 255.92 215.57 215.79 215.83 215.91 215.87 215.87 215.87 215.87 215.87 215.22 215.32 215.73 215.32 215.14 214.70 214.52 215.97 215.05 215.67 215.91 215.83 215.83 215.83 215.89 215.89 215.89 215.89 216.19 17.41 17.40 17.40 17.41 17.41 AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbinc Inserts and Without Boraflex ANP-2843(NP) Revision 1 Pale C-9 No Case name 48 49 50 51 52 53 54 55 56 57 58 59 60 61 62 63 64 65 66 67 68 69 70 71 72 73 74 75 76 77 78 79 80 81 82 83 84 85 86 87 88 89 90 91 92 93 94 95 96 97 rcon06 rcon07 rcon08 rcon09 rcon 10 rcon 11 rcon 12 rcon 13 rcon 14 rcon 15 rcon 16 rcon 17 rcon 18 rcon 19 rcon20 rcon2l rcon28 mdis0l mdis02 mdis03 mdis04 mdis05 mdis06 mdis07 mdis08 mdis09 mdisl0 mdisl 1 mdis12 mdis13 mdis14 mdis15 mdis16 mdis17 mdis18 mdis19 leuct022-02 leuct022-03 leuct024-01 leuct024-02 leuct025-01 leuct025-02 epri70b (PNL-31)epri70un (PNL-30)epri87b (PNL-33)epri87un (PNL-32)epri99b (PNL-35)epri99un (PNL-34)saxtnl04 (case 6)saxtn56b (case 3)Benchmark values SCALE 4.4a Calculated Values keff Texp kff Gcalc 1.0007 0.0006 0.9982 0.0007 1.0007 0.0006 0.9984 0.0006 1.0007 0.0006 1.0155 0.0008 1.0007 0.0006 0.9973 0.0007 1.0007 0.0006 0.9982 0.0008 1.0007 0.0006 0.9958 0.0007 1.0007 0.0006 0.9979 0.0007 1.0007 0.0006 0.9971 0.0006 1.0007 0.0006 0.9967 0.0007 1.0007 0.0006 0.9980 0.0006 1.0007 0.0006 0.9954 0.0006 1.0007 0.0006 0.9963 0.0007 1.0007 0.0006 0.9929 0.0007 1.0007 0.0006 0.9952 0.0007 1.0007 0.0006 0.9952 0.0007 1.0007 0.0006 0.9945 0.0007 1.0007 0.0006 0.9970 0.0008 1.0000 0.0014 0.9929 0.0008 1.0000 0.0014 0.9862 0.0009 1.0000 0.0014 0.9845 0.0009 1.0000 0.0014 0.9895 0.0008 1.0000 0.0014 0.9901 0.0009 1.0000 0.0014 1.0010 0.0008 1.0000 0.0014 0.9901 0.0009 1.0000 0.0014 0.9858 0.0008 1.0000 0.0014 0.9856 0.0009 1.0000 0.0014 0.9928 0.0009 1.0000 0.0014 1.0029 0.0009 1.0000 0.0014 1.0080 0.0008 1.0000 0.0014 0.9916 0.0009 1.0000 0.0014 0.9887 0.0008 1.0000 0.0014 0.9881 0.0010 1.0000 0.0014 1.0015 0.0008 1.0000 0.0014 0.9987 0.0008 1.0000 0.0014 0.9961 0.0008 1.0000 0.0014 0.9928 0.0009 1.0000 0.0046 1.0056 0.0013 1.0000 0.0036 1.0048 0.0013 1.0000 0.0054 0.9990 0.0015 1.0000 0.0040 1.0048 0.0014 1.0000 0.0041 0.9851 0.0014 1.0000 0.0044 0.9936 0.0013 1.0009 0.0047 0.9995 0.0016 1.0024 0.0060 0.9967 0.0015 1.0024 0.0024 1.0046 0.0013 EALF Enr (eV) wt%2 3 5 U 2.5119 2.46 1.6313 2.46 1.1134 2.46 1.4481 2.46 1.4623 2.46 1.5006 2.46 1.4942 2.46 1.4973 2.46 1.5185 2.46 1.5122 2.46 0.4182 2.46 0.4293 2.46 0.4354 2.46 0.4371 2.46 0.4367 2.46 0.4404 2.46 0.9984 2.46 0.2822 4.74 0.2641 4.74 0.2636 4.74 0.2513 4.74 0.2411 4.74 0.2292 4.74 0.2250 4.74 0.2493 4.74 0.2483 4.74 0.2221 4.74 0.2043 4.74 0.1946 4.74 0.1947 4.74 0.2299 4.74 0.2270 4.74 0.1905 4.74 0.1794 4.74 0.1747 4.74 0.1747 4.74 0.2920 9.83 0.1253 9.83 1.0568 9.83 0.1435 9.83 0.4401 7.41 0.2015 7.41 0.7631 -0.5648 0.2780 B (ppm)335 361 121 886 871 852 834 815 781 746 1156 1141 1123 1107 1093 1068 121 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 688 2 1090 1 767 2 0 337 H/X 17.41 17.43 17.43 44.81 44.81 44.79 44.81 44.81 44.79 44.79 118.47 118.47 118.44 118.44 118.44 118.44 17.44 137.61 137.61 137.61 137.61 137.61 137.61 137.61 137.61 137.61 137.61 137.61 137.61 137.61 137.61 137.61 137.61 137.61 137.61 137.61 80.00 151.00 41.00 128.00 66.30 106.10 146.15 146.20 308.83 308.99 445.41 445.57 473.11 95.24 1.0042 0.0031 1.0034 0.0013 .0.1894 1.0029 0.0027 1.0066 0.0009 0.1802 1.0038 0.0025 1.0088 0.0019 0.1353 1.0000 0.0023 1.0056 0.0017 0.1001 1.0000 0.0054 0.9980 0.0019 0.6523 AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page C-10 No Case name 98 saxtn792 (case 5)99 saxton52 (case 1)100 saxton56 (case 2)Benchmark values keff 1.0049 1.0028 1.0019 aexp 0.0027 0.0072 0.0059 SCALE 4.4a Calculated Values keff Gcalc 1.0027 0.0019 0.9987 0.0013 0.9997 0.0018 EALF (eV)0.1547 0.8878 0.5450 Enr wt%2 3 5 U B (ppm)0 0 0 H/X 249.70 73.86 95.29 In order to address situations in which the critical experiment being modeled was at other than a critical state (i.e., slightly super or subcritical), the calculated keff is normalized to the experimental kexp, using the following formula (Eq.9 from Reference C.2): knorm .. kcajc / kcxp In the following, the normalized values of the keff were used in the determination of the code bias and bias uncertainty. C.5 Trending Analysis The next step of the statistical methodology used to evaluate the code bias for the pool of experiments selected is to identify any trend in the bias. This is done by using the trending parameters presented in Table C.4.AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel ANP-2843(NP) Storage Pool Criticality Safety Analysis with Revision 1 Neutron Absorbing Inserts and Without Boraflex Page C-11 Table C.4 Trending Parameters Energy of the Average Lethargy causing Fission (EALF)Fuel Enrichment (wt% 2 3 5 U)Atom ratio of the moderator to fuel (H/X)Soluble Boron Concentration The first step in calculating the bias uncertainty limit is to apply regression-based methods to identify any trending of the calculated values of keff with the spectral and/or physical parameters. The trends show the results of systematic errors or bias inherent in the calculational method used to estimate criticality. For the critical benchmark experiments that were slightly super or subcritical, an adjustment to the keff value calculated with SCALE 4.4a (kcaic) was done as suggested in Reference C.2. This adjustment is done by normalizing the calculated (kcaic) value to the experimental value (kexp).This normalization does not affect the inherent bias in the calculation due to very small differences in keff. Unless otherwise mentioned, the normalized keff values (knorm) have been used in all subsequent calculations. Each subset of normalized keff values is first tested for trending against the spectral and/or physical parameters of interest (in this case, presented in Table C.4 above), using the built-in regression analysis tool from any general statistical software (e.g., Excel). Trending in this context is linear regression of unweighted calculated keff on the predictor variable(s) (spectral and/or physical parameters). In addition, the equations presented in Reference C.2 are also applied to check for a linear dependency in case of weighted keff, using as weight the factor I/ o7,2 as previously discussed. The linear regression fitted equation is in the form y(x) = a + bx, where y is the dependent variable (keff) and x is any of the predictor variables mentioned in Table C.4. The difference between the predicted y and actual value is known as the random error component (residuals). AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel ANP-2843(NP) Storage Pool Criticality Safety Analysis with Revision 1 Neutron Absorbing Inserts and Without Boraflex Page C-12 The final validity of each linear trend is checked using well-established indicators or goodness-of-fit tests concerning the regression parameters. As a first indicator, the coefficient of determination (r 2) that is available as a result of using linear regression statistics, can be used to evaluate the linear trending. It represents the proportion of the sum of squares of deviations of the y values about their mean that can be attributed to a linear relation between y and x.Another assessment of the adequacy of the linear model can be done by checking the goodness-of-fit against a null hypothesis on the slope (b) (Reference C.7, p. 371). The slope test requires calculating the test statistic "T" as in the following equation along with the corresponding statistical parameters (Reference C.7, p. 371).T=A iS~where,/?1 is the estimated slope of the fitted linear regression equation, i=I,n and, (n -2) Zty, 2 where, 5, is the estimated value using the regression equation.The test statistic is compared to the Student t-distribution (tM 2 ,n-2) with 95% confidence and n-2 degrees of freedom (Reference C.8, p.T-5), where n is the initial number of points in the subset.Given a null hypothesis Ho:p3=O, of "no statistically significant trend exists (slope is zero)", the hypothesis would be rejected if ITI > t,/2 ,n-2 .By only accepting linear trends that the data supports with 95% confidence, trends due to the randomness of the data are eliminated. A good indicator of this statistical process is evaluation of the P-value probability that gives a direct estimation of the probability of having linear trending due only to chance.The last step of the regression analysis is determining whether or not the final requirements of the simple linear regression model are satisfied. The error components (residuals) need to be normally distributed with mean zero, and also the residuals need to show a random scatter AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page C-13 about the center line (no pattern). These requirements were verified for the present calculation by applying an omnibus normality test (Reference C.8, p.372) on the residuals. The results of the trending parameter analysis for the criticality benchmark set (unweighted keff)are summarized in Table C.5.Table C.5 Summary of Trending Analysis Trend Goodness-of-Valid Parameter n Intercept Slope r 2 T tO.0 2 5 ,n-2 P-value fit Tests Trend EALF 100 0.9937 0.002 0.061 2.53 1.987 0.013 Not Passed No (residuals not normal and show a pattern-see Figure IC.5)Enrichment 90a 0.9911 0.0008 0.070 2.57 1.991 0.012 Not passed No (wt% 2 3 5 U) (residuals not normal and show a pattern-see Figure C.6)H/X 100 0.9952 -2.2E-06 0.001 -0.37 1.987 0.714 Not Passed No Boron in 100 0.9945 1.5E-06 0.009 0.95 1.987 0.345 Not passed No moderator (ppm)a Benchmark experiments with MOX fuel excluded.The results in Table C.5 show that there are no statistically significant or valid trends of keff with the trending parameters. An additional check was done by checking if there are any trends on the weighted data. The results of the regression analysis obtained using weighted keff (with the weight factor 1/o-,2 as previously discussed) show that the determination coefficient (r 2) has similar low values as in the above table, indicating very weak and statistically insignificant trends.Figures C.1 to C.4 show the distribution of the normalized keff values versus the trending parameters investigated. AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page C-14 1.02 1.01 1e 0.99 0.98 4 0.97 0.00 0.50 1.00 1.50 2.00 2.50 3.00 EALF (eV)Figure C.1 Distribution of keff Data versus EALF for the Selected Pool of Benchmark Experiments Figure C.2 Distribution of keff Data versus Enrichment (2 3 5 U) for the Selected Pool of Benchmark Experiments AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page C-15 1.02 1.01 0.99 0.98-0.97 0.00 50.00 100.00 150.00 200.00 250.00 300.00 350.00 400.00 450.00 500.00 H/X Figure C.3 Distribution of keff Data versus HIX for the Selected Pool of Benchmark Experiments .=*1.02 1.01 1 0.99 , 0.98 0.97 0.00 200.00 400.00 600.00 800.00 1000.00 1200.00 1400.00 Boron (ppm)Figure C.4 Distribution of keff Data versus Soluble Boron Concentration for the Selected Pool of Benchmark Experiments AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page C-16 4.0000 3.0000 2.0000 '-5;1.0000 0.0000" 0.0300 0)5000 1.0000 1.5-bOO 2.0000 2.5000 3.0 00 U) -1.0000-2.0000-3.0000 EALF (eV)Figure C.5 Plot of Standard Residuals for Regression Analysis with EALF as Trending Parameter 5.0000 4.0000 -3.0000 S2.0000 0 0 1.0000 C0.0000 9 Mo ca 0. 0 00 2.000 4.0000 : 6.0000 ,8. 0000 10.0000 12.(000-1.0000-2.0000-3.0000 Enrichment (wt %2 3 5 U)Figure C.6 Plot of Standard Residuals for Regression Analysis with Enrichment as Trending Parameter AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel ANP-2843(NP) Storage Pool Criticality Safety Analysis with Revision 1 Neutron Absorbing Inserts and Without Boraflex Page C-17 C.6 Bias and Bias Uncertainty For situations in which no significant trending in bias is identified, the statistical methodology presented in Reference C.2 suggests to first check the normality of the pool of keff data.Applying the Shapiro-Wilk test (Reference C.2) the null hypothesis of a normal distribution is not rejected. A visual inspection of the normal probability plot of the keff data shows that the pool of keff data for the selected benchmarks can be considered normally distributed. This situation allows the application of the weighted single-sided lower tolerance limit to determine the bias uncertainty (Reference C.2). First by determining the factor for 95%probability at the 95% confidence level (C95/95) and then multiplying it with the evaluated squared-root of the pooled variance, the uncertainty limit is determined. From Reference C.9, C95/95 for n equal to 100 is 1.927. The squared root of the pooled variance calculated using the formulas presented is: SP= 2+ 2 = (2.45212E-05+1.63005E-06) 0°5 =0.00511 Bias Uncertainty = C95/95"= 1.927

0.00511 = 0.00985 The bias is obtained using the formula that includes the weighted average of keff Bias = keff -1 = 0.99458 -1 = -0.00542 These represent the final results which can be used to evaluate the maximum keff and k 9 5/9 5 values in the criticality analysis of the spent fuel pool. Note that this bias will be applied as a positive penalty in the equation for computation of k 9 5/9 5.C.7 Area of Applicability A brief description of the spectral and physical parameters characterizing the set of selected benchmark experiments is provided in Table C.6.AREVA NP Inc.

AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page C-18 Table C.6 Range of Values of Key Parameters in Benchmark Experiments Parameter Range of Values Geometrical shape Heterogeneous lattices;Rectangular and hexagonal Fuel type U0 2 rods MOX fuel rods Enrichment (for U0 2 fuel) 2.46 to 9.83 wt % 2 3 5 U Lattice pitch 1.04 to 2.6416 cm H/X 17.4 to 473 EALF 0.11 to 2.51 eV Absorbers Soluble boron Boron in plates: Reflectors Water Stainless Steel Aluminum C.8 Bias Summary and Conclusions This evaluation considers a selected set of criticality benchmark experiments with enrichments ranging from about 2.5 to about 10 wt% 2 3 5 U and includes some experiments with MOX fuel rods. The results of the evaluation provide the following information relative to the SCALE4.4a bias: Bias = kIef -1 = 0.99458 -1 = -0.00542*Note that this bias will be applied as a positive penalty in the equation for computation of k 9 5 1 9 5: Bias Uncertainty = 095/95* sP = 1.927

0.00511i 0.00985 The bias and its uncertainty (95/95 weighted single-sided tolerance limit) was obtained applying the appropriate steps of the statistical methodology presented in NUREG 6698 (Reference C.2)taking into account the possible trending of keff with various spectral and/or physical parameters.
This will be applied as biasm = 0.00542 in Section 6.6.t SP will be applied as am = 0.00511 in Section 6.6. This is because the one sided tolerance multiplier is applied to the combined uncertainties in Section 6.6.AREVA NP Inc.

AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel ANP-2843(NP) Storage Pool Criticality Safety Analysis with Revision 1 Neutron Absorbing Inserts and Without Boraflex Page C-19 C.9 References C.1 Nuclear Energy Agency, "International Handbook of Evaluated Criticality Safety Benchmark Experiments," NEA/NSC/DOC(95)03, Nuclear Energy Agency, Organization for Co-operation and Development, 2008.C.2 Nuclear Regulatory Commission, "Guide for Validation of Nuclear Criticality Safety Calculational Methodology", NUREG/CR-6698, January 2001.C.3 Bierman, S.R., Durst, B.M., Clayton, E.D., "Critical Separation Between Subcritical Clusters of 4.29 Wt% 2 3 5 U Rods in Water With Fixed Neutron Poisons," Battelle Pacific Northwest Laboratories, NUREG/CR-0073(PNL-2615). C.4 Baldwin, M.N., et.al., "Critical Experiments Supporting Close Proximity Water Storage Of Power Reactor Fuel," BAW-1484-7, July 1979.C.5 Hoovler, G.S., et.al., "Critical Experiments Supporting Underwater Storage of Tightly Packed Configurations of Spent Fuel Pins," BAW-1645-4, November 1981.C.6 "Dissolution and Storage Experimental Program with U(4.75)0 2 Rods," Transactions of the American Nuclear Society, Vol. 33, pg. 362.C.7 Rosenkrantz W.A., Introduction to Probability and Statistics for Scientists and Engineers, The McGraw-Hill, New York, NY, 1989.C.8 D'Agostino, R.B. and Stephens, M.A., Goodness-of-fit Techniques. Statistics, Textbooks and Monographs, Volume 68, New York, New York, 1986.C.9 Owen, D.B., Handbook of Statistical Tables, Addison-Wesley, Reading, MA.AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel ANP-2843(NP) Storage Pool Criticality Safety Analysis with Revision 1 Neutron Absorbing Inserts and Without Boraflex Page D-1 Appendix D CASMO-4 Benchmarking for In-Rack Modeling-D.1 Introduction The purpose of this Appendix is to provide qualification of the CASMO-4 code for use in the evaluation of the LaSalle Unit 2 Spent Fuel Pool with NETCO-SNAP-IN inserts. While the CASMO-4 code is not being used for the actual criticality calculation methodology, it is used for the selection of peak reactivity lattices and the determination of manufacturing uncertainties which have a depletion dependence. This evaluation is performed to address the guidance of References D.1 and D.2. The format and presentation follows the sample format presented in Section 6 of Reference D.2.D.2 Code System CASMO-4 is a multi-group, two-dimensional transport theory code with an in-rack geometry option where typical storage rack geometries can be defined on an infinite lattice basis. This code is used for fuel depletion and relative reactivity comparisons in a manner that is consistent with AREVA's NRC approved CASMO-4 / MICROBURN-B2 methodology (Reference D.3). The library files used in the evaluation are the standard CASMO-4 70 group library based on ENDFB-IV. The CASMO-4 computer code and data library are controlled by AREVA procedures and the version used in this analysis meets the requirements of Reference D.3. The CASMO-4 program is run on AREVA's HP-UX1 1 engineering workstations. D.3 Benchmarking Methodology Since the CASMO-4 code is a two-dimensional code that models the storage rack in an infinite array, it cannot be used to provide a stand-alone benchmark of finite criticality experiments. Consequently, the evaluation in this appendix takes a different approach -it provides a code to code comparison of the CASMO-4 code to the SCALE 4.4a KENO code. Benchmarking of the KENO code to criticality experiments was previously described in Appendix C.AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel ANP-2843(NP) Storage Pool Criticality Safety Analysis with Revision 1 Neutron Absorbing Inserts and Without Boraflex Page D-2 The benchmarking of the CASMO-4 code in this Appendix is performed in two steps to demonstrate its acceptability for the two different ways that CASMO-4 is used in the LaSalle analysis.Identify the relative reactivity of a lattice with the use of the storage rack geometry option. This is addressed by determining the CASMO-4 uncertainty relative to KENO by comparison of calculated k-infinities from the two codes.Evaluate relative changes in reactivity associated with changes in manufacturing tolerances. For this evaluation, the differential k-infinities from the two codes are compared based upon the same input perturbations. These different approaches are described in more detail in the following sections.In addition to benchmarking against KENO, the CASMO-4 depletion uncertainty is established based on Reference D.3.D.3.1 CASMO-4 Uncertainty for Absolute k-infinite Relative to KENO The approach that is taken for the benchmarking of the in-rack CASMO-4 model is to perform a series of calculations with varied enrichments, geometries, and temperatures. The results of the CASMO-4 calculations are then compared to KENO results for the same configurations. The validation guidance of NUREG/CR-6698 (Reference D.2) is followed to determine a code uncertainty for CASMO-4 relative to KENO. The KENO calculations are treated as the critical experiments in the validation process. The validation includes ATRIUM-10 top and bottom lattices as well as ATRIUM-9 lattices.D.3.2 CASMO-4 Uncertainty for Ak-infinite Relative to KENO The capability of the CASMO-4 code to predict the change in reactivity associated with a perturbation of fuel parameters is demonstrated by comparison of Ak values obtained with KENO to those obtained with CASMO-4. The approach taken is to evaluate small perturbations in reactivity by varying the enrichment relative to a base case. The same cases used in the evaluation of the uncertainty of the absolute multiplication factor are used in this evaluation. The Ak values will be determined for both KENO and CASMO-4 for enrichment perturbations from the reference case.AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel ANP-2843(NP) Storage Pool Criticality Safety Analysis with Revision 1 Neutron Absorbing Inserts and Without Boraflex Page D-3 The Ak values are compared between the two codes and a statistical evaluation similar to that identified in Reference D.2 is used to establish an uncertainty for the determination of Ak values with CASMO-4 relative to KENO.D.3.3 CASMO-4 Depletion Uncertainty The CASMO-4 depletion uncertainty is derived from the AREVA licensing topical report based on the extensive benchmarking that is documented within Reference D.3. Comparisons against critical experiments were performed by Studsvik with results reported in Table 2.1 of AREVA's CASMO-4/MICROBURN-B2 licensing topical report (Reference D.3). In addition, the beginning of cycle cold critical calculations reported in Table 2.2 of this same licensing topical report also provide comparisons to critical data. Results of these comparisons indicate that CASMO-4 results will have a standard deviation of [ ] Ak (Table 2.1 of Reference D.3) without depletion and a standard deviation of [ ] Ak (Table 2.2 of Reference D.3) when the majority of assemblies have been depleted*. In addition to depletion effects, the [ ] Ak standard deviation from Reference D.3 also includes manufacturing and measurement uncertainties. Since it is difficult to separate these uncertainties, this entire value ([ ] Ak) will be used for the CASMO-4 depletion uncertainty when using the discrete void history levels from Reference D.3.D.4 Experiment Descriptions As noted, KENO calculations are used as the reference experiments. The evaluations are based on the LaSalle Unit 2 Spent Fuel Pool with NETCO-SNAP-IN inserts. The validation is performed using both bottom and top ATRIUM-10 and ATRIUM-9 lattice geometries within the LaSalle Unit 2 Spent Fuel Pool with NETCO-SNAP-IN inserts. Enrichment is varied in 0.05 increments above and below an assumed base enrichment level up to maximum delta of 0.25.The maximum peak reactivity of the fuel manufactured for LaSalle in the given geometry is represented within the range of enrichments evaluated. The calculations are reported for 40C, 20 0 C and 1000C (2771K, 2931K and 3731K).The uncertainty of cold critical benchmarks effectively includes a depletion uncertainty since the majority of the bundles in the core are exposed. It is noted, that a cold in-sequence critical has significant similarities to an in-rack calculation since the majority of the control blades remain inserted effectively surrounding the majority of the fuel with a strong neutron absorber on two sides.AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page D-4 The minimum, base, and maximum enrichments for the ATRIUM-10 bottom (AlOB), the ATRIUM-10 top (A1OT) and the ATRIUM-9 (AT9) lattices are: I The fuel assembly data, rack geometry, and NETCO-SNAP-IN insert are the same as those for the LaSalle Unit 2 Spent Fuel Pool configuration. D.5 Analysis of Validation Results D.5.1 CASMO-4 Uncertainty for Absolute k-effective Relative to KENO The calculated multiplication factors from KENO and CASMO were tabulated. The O'keno terms are taken from each individual KENO calculation and the 0 casmo terms are set to the CASMO-4 convergence criteria for the individual case. (Use of the CASMO convergence is consistent with footnote 1 on page 6 of Reference D.2.) A combined uncertainty atotwas determined consistent with equation 3 of Reference D.2.t~ 02 2 Otot = 'keno + U"c..smo The tabulated results are provided in Table D.1. The geometry is identified as either AlOB (bottom lattice), AIOT (top lattice), or AT9 (ATRIUM-9) along with the temperature and enrichment variations. AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page D-5 Table D.1 CASMO4 and KENO Validation Case Information I I AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page D-6 Table D.1 CASMO4 and KENO Validation Case Information (Continued) I I AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page D-7 Table D.1 CASMO4 and KENO Validation Case Information (Continued) I I Since this is a comparison between two codes, the differences of the calculated values for the multiplication factor are determined. The results of the difference along with the components used in the statistical evaluation are provided in Table D.2.AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbinal Inserts and Without Boraflex ANP-2843(NP) Revision 1 Paae D-8 Table D.2 CASMO -KENO Difference and Statistical Parameters I I Ak is kCASMO -kKENO AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbina Inserts and Without Boraflex ANP-2843(NP) Revision 1 Paae D-9 Table D.2 CASMO -KENO Difference and Statistical Parameters (Continued) I I Ak is kCASMO -kKENO AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page D-10 Table D.2 CASMO -KENO Difference and Statistical Parameters (Continued) The weighted average difference (Akbar), the variance about the mean S2, and the average total uncertainty G 2 are calculated using the weighting factor 1/Vat 2.The square root of the pooled variance is determined per Equation 7 of Reference D.2 Ak is kCASMO -kKENO AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page D-11 Sp = 2+ O 2 Weighted mean difference Average total uncertainty Variance about mean Square root of pooled variance Akbar sz Sp Sp[I L[Ref D.2 Eq 6 1 Ref D.2 Eq 5 Ref D.2 Eq 4 Ref D.2 Eq 7 The CASMO-4 bias relative to KENO is[ I.]. The bias uncertainty value is rounded up to AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel ANP-2843(NP) Storage Pool Criticality Safety Analysis with Revision 1 Neutron Absorbing Inserts and Without Boraflex Page D-12 Normality test Normality tests were performed on the combined data and the results were somewhat indeterminate but indicated potential non-normality. The data was then subdivided by temperature which is consistent with the use of CASMO-4 in comparing lattice results at the same temperature. In this comparison each temperature data set was determined to be a normal distribution. A single uncertainty for the combined data set is conservatively reported rather than individual temperature dependent uncertainty values.Since this uncertainty value is only used to demonstrate that the CASMO-4 code can select the most reactive lattices for a given temperature, a 95/95 confidence multiplier is not determined. Data Trendinq No specific trending of the code bias was completed since CASMO-4 is not used directly for the determination of the absolute value of the multiplication factor. It is noted that the agreement is better at 4 0C than 100 0C.Area of Applicability The fuel and rack geometry as well as fuel enrichment were evaluated consistent with the LaSalle Unit 2 spent fuel pool. Therefore the area of applicability is specific to the LaSalle Unit 2 spent fuel pool with inserts.D.5.2 CASMO-4 Uncertainty for Ak-effective The actual KENO and CASMO calculations used in this evaluation are those used in Section D.5.1. In this evaluation, the relative reactivity change is evaluated by taking the delta with respect to the initial reference reactivity. A difference is then determined between the Ak values obtained with KENO and the Ak values obtained with CASMO-4 for the same perturbation. AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page D-13 Table D.3 Lattice Evaluations at 4VC I I AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbina Inserts and Without Boraflex ANP-2843(NP) Revision 1 Pace D-14 Table D.4 Lattice Evaluations at 20 0 C I I AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbina Inserts and Without Boraflex ANP-2843(NP) Revision 1 Paae D-15 Table D.5 Lattice Evaluations at 100°C I I The average difference between the Ak values was I I.] with a standard deviation of AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel ANP-2843(NP) Storage Pool Criticality Safety Analysis with Revision 1 Neutron Absorbing Inserts and Without Boraflex Page D-16 The Shapiro-Wilk data normality test and the Anderson-Darling goodness of fit for normality (see section 9.5.4.1 of Reference D.4) were performed on the Ak comparisons. Based on the test results and a visual inspection of the data, it is considered normally distributed. For the data sample of 50 the single sided tolerance factor is 2.065 from Table 2.1 of Reference D.2. This is conservatively applied for 90 data samples.Therefore, the 95/95 uncertainty is Data Trendinq A code bias is not used in the evaluation of incremental reactivity. Therefore, trending of the bias was not completed. Area of Applicability The fuel and rack geometry as well as fuel enrichment were evaluated consistent with the LaSalle Unit 2 spent fuel pool. Therefore the area of applicability is specific to the LaSalle Unit 2 spent fuel pool with inserts.AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbing Inserts and Without Boraflex ANP-2843(NP) Revision 1 Page D-17 D.6 Total CASMO-4 Uncertainty When applied on a differential basis a Ak predicted by CASMO-4 agrees with the KENO V.a based Ak with an uncertainty less than [ ] Ak, (see Section D.5). This can be combined with the [ ] Ak depletion uncertainty discussed in Section D.3.3 to obtain the total CASMO-4 uncertainty. A 95/95 uncertainty result is also obtained by multiplying these uncertainties by an appropriate multiplier. Since these values are independent they will be combined using the square root of the sum of the squares as shown in the following table. This process results in a total CASMO-4 uncertainty value of less than 0.007* Ak.Uncertainty Value a 95/95 Multiplier 95/95 Uncertainty Depletion [ ] 2.0 [Calculational (Ak based) 2.065 [Combined [An alternate approach for determining the reactivity worth of the uncertainty in the fuel depletion calculation is discussed in Section 5.A.5 of Reference D.1. "In the absence of any other determination of the depletion uncertainty, an uncertainty equal to 5% of the reactivity decrement to the burnup of interest is an acceptable assumption." While this section of Reference D.1 explicitly addresses analyses that credit reactivity depletion due to fuel burnup (i.e. burnup credit), recent discussions with the NRC indicate that 5% of the reactivity increment (BOL to peak reactivity) would be an acceptable representation of the depletion uncertainty to peak reactivity. Based on this information, 5% of the reactivity increment from BOL to peak reactivity was determined for the three reference bounding lattices. [] Therefore, the uncertainty of a single assembly made up of these lattices will not differ significantly from the 0.007 Ak determined here.AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel ANP-2843(NP) Storage Pool Criticality Safety Analysis with Revision 1 Neutron Absorbing Inserts and Without Boraflex Page D-18 D.7 Conclusions A code bias uncertainty of [ ] was determined for CASMO-4 relative to KENO V.a in the determination of the absolute value of the multiplication factor. Based on this, it is demonstrated that the CASMO-4 code can be used for the characterization of the in-rack reactivity of fuel designs in the LaSalle Unit 2 spent fuel pool.A standard deviation of [ ] was established for determining Ak with CASMO-4 relative to the Ak determined with KENO V.a. A 95/95 confidence multiplier of 2.065 is applicable for this uncertainty. The evaluation of the ATRIUM-9, ATRIUM-10 bottom, and ATRIUM-10 top lattices demonstrated that there is no specific fuel geometry dependence relative to the use of CASMO-4 with respect to evaluating the in-rack reactivity. The 0.01 Ak adder used when defining the REBOL lattices conservatively bounds the CASMO-4 uncertainty. Consequently, no CASMO bias or uncertainty is required in the final k 9 5/9 5 calculation. AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel ANP-2843(NP) Storage Pool Criticality Safety Analysis with Revision 1 Neutron Absorbing Inserts and Without Boraflex Page D-19 D.8 References D.1 Memorandum L. Kopp to T. Collins, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," NRC, August 19, 1998. (NRC -ADAMS Accession Number ML072710248) D.2 NUREG/CR-6698, "Guide for Validation of Nuclear Criticality Safety Calculational Methodology," USNRC, January 2001.D.3 EMF-2158(P)(A) Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2, Siemens Power Corporation, October 1999.D.4 MIL-HDBK-5J, "Metallic Materials and Elements for Aerospace Vehicle Structures", Department of Defense, January 2003.AREVA NP Inc. AREVA NP LaSalle Unit 2 Nuclear Power Station Spent Fuel Storage Pool Criticality Safety Analysis with Neutron Absorbina Inserts and Without Boraflex ANP-2843(NP) Revision 1 Distribution Controlled Distribution Richland RJ RE R SW DP CD CM EE AW PD DeMartino Fowles Fundak Jones Jordheim Manning Powers Riley Will Wimpy AREVA NP Inc. ATTACHMENT 7 Summary of Regulatory Commitments The following list identifies those actions committed to by Exelon Generation Company, LLC, (EGC) in this submittal. Any other actions discussed in the submittal represent intended or planned actions by EGC, are described only for information, and are not regulatory commitments. COMMITMENT TYPE ONE-TIME PROGRAM-ACTION MATIC COMMITTED DATE COMMITMENT OR "OUTAGE" (YESINO) (YESINO)The ATRIUM-10 fuel assembly Upon implementation of No Yes design limitations will be the proposed change incorporated in reload design documents and SFP criticality compliance procedures. Additionally, the design limitations will be reflected in Sections 9.1.2.1 and 9.1.2.2 of the LaSalle County Station (LSCS) Updated Final Safety Analysis Report (UFSAR).The Boraflex monitoring program Complete No Yes will continue to be maintained for as long as EGC continues to credit Boraflex for criticality control, regardless of the implementation of NETCO-SNAP-IN rack inserts.The rack inserts will be installed in Prior to crediting the Yes No stages, with each stage of neutron absorption installation resulting in the use of a capabilities of the rack insert in all the spent fuel NETCO-SNAP-IN rack storage rack cells of a given inserts for each individual spent fuel storage rack individual Unit 2 spent and all the cells of the first row and fuel storage rack first column of adjoining spent fuel storage racks, such that all sides of the fuel assemblies within the spent fuel storage rack are adjacent to a face of the rack insert's wing.Page 1 ATTACHMENT 7 Summary of Regulatory Commitments COMMITMENT TYPE ONE-TIME PROGRAM-ACTION MATIC COMMITTED DATE COMMITMENT OR "OUTAGE" (YES/NO) (YES/NO)EGC will implement the Rio Tinto Upon implementation of No Yes Alcan Composite Surveillance the proposed change Program as described in Section 3.9 of Attachment 1 to ensure that the performance requirements of the Rio Tinto Alcan composite in the NETCO-SNAP-IN rack inserts are met over the lifetime of the spent fuel storage racks with the rack inserts installed. A description of the program will be added to the LSCS UFSAR upon implementation of the proposed change..1 Page 2}}

v t e Letter Sequence Request
TAC:ME2376, (Approved, Closed)
Initiation Request, Request, Request, Request, Request, Request, Request Acceptance... Administration Withholding Request Acceptance Meeting, Meeting, Meeting, Meeting, Meeting, Meeting Questions 26 April 2010 3 May 2010 23 June 2010 19 November 2010 10 December 2010 Answers 18 April 2011 Results Approval Other: ML101650229, ML110050255, ML110180331, ML110460617, RS-11-109, Completion of License Condition 2.C.(30) Associated with Unit 2 Spent Fuel Pool Storage Racks, RS-11-170, Completion of License Condition 2.C.(31) Associated with Unit 2 Spent Fuel Pool Storage Racks
MONTHYEAR2010-04-26 [Table View]

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